Analysis of temperature and stress distribution of superheater tubes after attemperation or sootblower activation

International Nuclear Information System (INIS)

Madejski, Paweł; Taler, Dawid

2013-01-01

Highlights: • The CFD simulation was used to calculate 3D steam and tube wall temperature distributions in the platen superheater. • The CFD results can be used in design of superheaters made of tubes with complex cross-section. • The CFD analysis enables the proper selection of the steel grade. • The transient temperature and stress distributions were calculated using Finite Volume Method. • The detailed analysis prevents superheater tubes from excessive stresses during sootblower or attemperator activation. - Abstract: Superheaters are characterized by high metal temperatures due to higher steam temperature and low heat transfer coefficients on the tube inner surfaces. Superheaters have especially difficult operating conditions, particularly during attemperator and sootblower activations, when temperature and steam flow rate as well as tube wall temperature change with time. A detailed thermo-mechanical analysis of the superheater tubes makes it possible to identify the cause of premature high-temperature failures and aids greatly in the changes in tubing arrangement and improving start-up technology. This paper presents a thermal and strength analysis of a tube “double omega”, used in the steam superheaters in CFB boilers

Bonus schemes and trading activity

NARCIS (Netherlands)

Pikulina, E.S.; Renneboog, L.D.R.; ter Horst, J.R.; Tobler, P.N.

2014-01-01

Little is known about how different bonus schemes affect traders' propensity to trade and which bonus schemes improve traders' performance. We study the effects of linear versus threshold bonus schemes on traders' behavior. Traders buy and sell shares in an experimental stock market on the basis of

Single Temperature Sensor Superheat Control Using a Novel Maximum Slope-seeking Method

DEFF Research Database (Denmark)

Vinther, Kasper; Rasmussen, Henrik; Izadi-Zamanabadi, Roozbeh

2013-01-01

Superheating of refrigerant in the evaporator is an important aspect of safe operation of refrigeration systems. The level of superheat is typically controlled by adjusting the flow of refrigerant using an electronic expansion valve, where the superheat is calculated using measurements from...

Nonlinear Superheat and Evaporation Temperature Control of a Refrigeration Plant

DEFF Research Database (Denmark)

Rasmussen, Henrik; Thybo, Claus; Larsen, Lars F. S.

2006-01-01

This paper proposes novel control of the superheat of the evaporator in a refrigeration system. A new model of the evaporator is developed and based on this model the superheat is transferred to a referred variable. It is shown that control of this variable leads to a linear system independent...... of the working point. The model also gives a method for control of the evaporation temperature. The proposed method is validated by experimental results....

Advanced steam cycles for light water reactors. Final report

International Nuclear Information System (INIS)

Mitchell, R.C.

1975-07-01

An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

Remaining Life Estimation Of Secondary Superheater Outlet On Industrial Electrical Boiler

International Nuclear Information System (INIS)

Soedardjo; Andryansyah; Arhatari, B.D.; Natsir, Muhammad; Triyadi, Ari; Farokhi

2001-01-01

Remaining life estimation of secondary superheater header outlet (SSHO) on industrial electrical boiler has been carried out. Estimation conducted by the observation of microstructure cavitation development based on Neubauer and Wedel theory. The result is available for isolated cavitation development present yet. That Secondary Superheater Outlet component is in good condition after 14 years operated and predicted could be operated for 36 years again

Oxidation rate in ferritic superheater materials

International Nuclear Information System (INIS)

Falk, I.

1992-05-01

On the steam side of superheater tubes, compact oxide layers are formed which have a tendency to crack and flake off (exfoliate). Oxide particles then travel with the steam and can give rise to erosion damage in valves and on turbine blades. In an evaluation of conditions in superheater tubes from Swedish power boilers, it was found that the exfoliation frequency for one material quality (SS 2218) was greater than for other qualities. Against this background, a literature study has been carried out in order to determine which mechanisms govern the build-up of oxide and the exfoliation phenomenon. The study reveals that the oxide morphology is similar on all ferritic steels with Cr contents up to 5%. and that the oxide properties can therefore be expected to be similar. The reason why the exfoliation frequency is greater for tubes of SS 2218 is probably that the tubes have been exposed to higher temperatures. SS 2218 (2.25 Cr) is normally used in a higher temperature range which is accompanied by improved strength data as compared with SS 2216 (1 Cr). The principal cause of the exfoliation is said to be stresses which arise in the oxide during the cooling-down process associated with shutdowns. The stresses give rise to longitudinal cracks in the oxide, and are formed as a result of differences in thermal expansion between the oxide and the tube material. In addition, accounts are presented of oxidation constants and growth velocities, and thickness and running time. These data constitute a valuable basis for practical estimates of the operating temperature in routine checks and investigations into damage in superheater tubes. (au)

Lifetime evaluation of superheater tubes exposed to steam oxidation, high temperature corrosion and creep

Energy Technology Data Exchange (ETDEWEB)

Henriksen, N [Elsamprojekt A/S, Faelleskemikerne, Fredericia (Denmark); Hede Larsen, O; Blum, R [I/S Fynsvaerket, Faelleskemikerne, Odense (Denmark)

1996-12-01

Advanced fossil fired plants operating at high steam temperatures require careful design of the superheaters. The German TRD design code normally used in Denmark is not precise enough for the design of superheaters with long lifetimes. The authors have developed a computer program to be used in the evaluation of superheater tube lifetime based on input related to tube dimensions, material, pressure, steam temperature, mass flux, heat flux and estimated corrosion rates. The program is described in the paper. As far as practically feasible, the model seems to give a true picture of the reality. For superheaters exposed to high heat fluxes or low internal heat transfer coefficients as is the case for superheaters located in fluidized bed environments or radiant environments, the program has been extremely useful for evaluation of surface temperature, oxide formation and lifetime. The total uncertainty of the method is mainly influenced by the uncertainty of the determination of the corrosion rate. More precise models describing the corrosion rate as a function of tube surface temperature, fuel parameters and boiler parameters need to be developed. (au) 21 refs.

DYNAM, Once Through Boiling Flow with Steam Superheat, Laplace Transformation

International Nuclear Information System (INIS)

Schlueter, G.; Efferding, L.E.

1973-01-01

1 - Description of problem or function: DYNAM performs a dynamic analysis of once-through boiling flow oscillations with steam superheat. The model describing the superheat regime (single- phase, variable density fluid) for subcritical pressure operation is also applicable to the study of once-through operation using supercritical pressure water. 2 - Method of solution: Linearized partial differential conservation equations are solved using Laplace transformation of the temporal terms and integration of the spatial variations. DYNAM is written in complex variable notation. 3 - Restrictions on the complexity of the problem - Maxima of: 30 intervals used to describe the power distribution in the non-boiling and boiling regions, 29 boiling nodes, 7 intervals and corresponding friction multipliers read in per case, 14 exit qualities read in per case, 40 superheat nodes, 10 coefficients read in for the phi 2 vs, x-polynomial fit, 48 frequencies at which open-loop frequency response is desired, 48 frequencies at which signal output is desired

Thermomechanical CSM analysis of a superheater tube in transient state

Science.gov (United States)

Taler, Dawid; Madejski, Paweł

2011-12-01

The paper presents a thermomechanical computational solid mechanics analysis (CSM) of a pipe "double omega", used in the steam superheaters in circulating fluidized bed (CFB) boilers. The complex cross-section shape of the "double omega" tubes requires more precise analysis in order to prevent from failure as a result of the excessive temperature and thermal stresses. The results have been obtained using the finite volume method for transient state of superheater. The calculation was carried out for the section of pipe made of low-alloy steel.

High-Temperature Graphitization Failure of Primary Superheater Tube

Science.gov (United States)

Ghosh, D.; Ray, S.; Roy, H.; Mandal, N.; Shukla, A. K.

2015-12-01

Failure of boiler tubes is the main cause of unit outages of the plant, which further affects the reliability, availability and safety of the unit. So failure analysis of boiler tubes is absolutely essential to predict the root cause of the failure and the steps are taken for future remedial action to prevent the failure in near future. This paper investigates the probable cause/causes of failure of the primary superheater tube in a thermal power plant boiler. Visual inspection, dimensional measurement, chemical analysis, metallographic examination and hardness measurement are conducted as the part of the investigative studies. Apart from these tests, mechanical testing and fractographic analysis are also conducted as supplements. Finally, it is concluded that the superheater tube is failed due to graphitization for prolonged exposure of the tube at higher temperature.

On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

International Nuclear Information System (INIS)

Malkis, V.A.; Lokshin, V.A.

1983-01-01

Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

Designing a Better Navy Aviation Retention Bonus

Science.gov (United States)

2017-03-01

limitations of a one-size-fits-all bonus and adopted a bonus system that rewarded good performance and incentivized officers to commit to the bonus...According to Kuhn and Yockey, the performance evaluation structure that best lends itself to a bonus system that offers a high-risk high- reward option and...performers sooner than if they waited for O4 or Department Head boards to screen them out. 47 4. Increased Motivation and Effort Under this new system

Influence of Superheated Steam Temperature Regulation Quality on Service Life of Boiler Steam Super-Heater Metal

Directory of Open Access Journals (Sweden)

G. T. Kulakov

2009-01-01

Full Text Available The paper investigates influence of change in quality of superheated steam temperature regulations on service life of super-heater metal. А dependence between metal service life and dispersion value for different steel grades has been determined in the paper. Numerical values pertaining to increase of super-heater metal service life in case of transferring from manual regulation to standard system of automatic regulation (SAR have been determined and in case of transferring from standard SAR to improved SAR. The analysis of tabular data and plotted dependencies makes it possible to conclude that any change in conditions of convection super-heater metal work due to better quality of the regulation leads to essential increase of time period which is left till the completion of the service life of a super-heater heating surface.

High-Speed Imaging of Explosive Droplet Boiling at the Superheat Limit

Science.gov (United States)

Ferris, F. Robert; Hermanson, Jim; Asadollahi, Arash; Esmaeeli, Asghar

2017-11-01

The explosive boiling processes of droplets of diethyl ether (1-2 mm in diameter) at the superheat limit were examined both experimentally and computationally. Experimentally, droplet explosion was studied using a heated bubble column to bring the test droplet to the superheat limit. The droplet fluid was diethyl ether (superheat limit 147 C at 1 bar) with immiscible glycerol employed as the heated host fluid. Tests were carried out at pressures between 0.5 and 4 bar absolute. The pressure rise associated with the explosive boiling event was captured using a piezoelectric quartz pressure transducer with a 1 MHz DAQ system. High-speed imaging of the interfacial behavior during explosive boiling was performed using a Phantom v12.1 camera at a frame rate of up to one million frames per second with the droplets illuminated by diffuse back-lighting. The imaging reveals features of the Rayleigh-Taylor instability at the vapor-liquid interface resulting from the unstable boiling process. Computationally, Direct Numerical Simulations are performed at Southern Illinois University Carbondale to compliment the experimental tests. NSF Award Number 1511152.

Composite electric generator equipped with steam generator for heating reactor coolant

International Nuclear Information System (INIS)

Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

1997-01-01

The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

Alkali chloride induced corrosion of superheaters under biomass firing conditions: Improved insights from laboratory scale studies

DEFF Research Database (Denmark)

Okoro, Sunday Chukwudi; Montgomery, Melanie; Jappe Frandsen, Flemming

2015-01-01

One of the major operational challenges experienced by power plants firing biomass is the high corrosion rate of superheaters. This limits the outlet steam temperature of the superheaters and consequently, the efficiency of the power plants. The high corrosion rates have been attributed to the fo......One of the major operational challenges experienced by power plants firing biomass is the high corrosion rate of superheaters. This limits the outlet steam temperature of the superheaters and consequently, the efficiency of the power plants. The high corrosion rates have been attributed......, [1–3]). However, complete understanding of the corrosion mechanism under biomass-firing conditions has not yet been achieved. This is attributed partly to the complex nature of the corrosion process since there are many species produced from fuel combustion which can interact with one another...... and the steel surface. Many studies have focused on specific parameters such as, deposit composition (KCl, K2SO4, K2CO3, etc.) or gas species such as HCl, SO2, H2O [4–6], however, more research is necessary to understand the interaction of deposits and gas mixtures with each other and metallic superheater...

Prosocial bonuses increase employee satisfaction and team performance.

Directory of Open Access Journals (Sweden)

Lalin Anik

Full Text Available In three field studies, we explore the impact of providing employees and teammates with prosocial bonuses, a novel type of bonus spent on others rather than on oneself. In Experiment 1, we show that prosocial bonuses in the form of donations to charity lead to happier and more satisfied employees at an Australian bank. In Experiments 2a and 2b, we show that prosocial bonuses in the form of expenditures on teammates lead to better performance in both sports teams in Canada and pharmaceutical sales teams in Belgium. These results suggest that a minor adjustment to employee bonuses--shifting the focus from the self to others--can produce measurable benefits for employees and organizations.

Prosocial Bonuses Increase Employee Satisfaction and Team Performance

Science.gov (United States)

Anik, Lalin; Aknin, Lara B.; Norton, Michael I.; Dunn, Elizabeth W.; Quoidbach, Jordi

2013-01-01

In three field studies, we explore the impact of providing employees and teammates with prosocial bonuses, a novel type of bonus spent on others rather than on oneself. In Experiment 1, we show that prosocial bonuses in the form of donations to charity lead to happier and more satisfied employees at an Australian bank. In Experiments 2a and 2b, we show that prosocial bonuses in the form of expenditures on teammates lead to better performance in both sports teams in Canada and pharmaceutical sales teams in Belgium. These results suggest that a minor adjustment to employee bonuses – shifting the focus from the self to others – can produce measurable benefits for employees and organizations. PMID:24058691

Contamination prevention of superheaters and reheaters during initial startup and operation

International Nuclear Information System (INIS)

Gabrielli, F.; Sylvester, W.R.; Thimot, G.W.

1976-01-01

The general precautions that should be taken to minimize the potential for harmful contamination or oxygen corrosion of power plant superheaters and reheaters during the period from field storage through operation are discussed and summarized. Present boiler industry start-up and operating practices intended to minimize the introduction of solids to the superheater are, as proven by experience, adequate to avoid contamination-related problems. No basic changes to general industry practice are necessary. What is needed, however, is a continuing awareness of the potential for contamination-related problems so that in the specific application of these practices all likely sources of contamination will be considered

Phase Identification and Internal Stress Analysis of Steamside Oxides on Plant Exposed Superheater Tubes

DEFF Research Database (Denmark)

Pantleon, Karen; Montgomery, Melanie

2012-01-01

During long-term, high-temperature exposure of superheater tubes in thermal power plants, various oxides are formed on the inner side (steamside) of the tubes, and oxide spallation is a serious problem for the power plant industry. Most often, oxidation in a steam atmosphere is investigated...... in laboratory experiments just mimicking the actual conditions in the power plant for simplified samples. On real plant-exposed superheater tubes, the steamside oxides are solely investigated microscopically. The feasibility of X-ray diffraction for the characterization of steamside oxidation on real plant......-exposed superheater tubes was proven in the current work; the challenges for depth-resolved phase analysis and phase-specific residual stress analysis at the inner side of the tubes with concave surface curvature are discussed. Essential differences between the steamside oxides formed on two different steels...

Survey of the current state of knowledge of incipient boiling superheat in sodium

International Nuclear Information System (INIS)

Greer, B.

1979-01-01

Superheat data obtained by various investigators indicate that many parameters affect this phenomenon. Controlling parameters appear to be inert gas concentration, oxide concentration, system pressure, pressure-temperature history, rate of temperature rise, heat flux, flow rate, operating time on the system, surface conditions, and radiation. Of these, the two believed most influential in controlling incipient boiling superheat are the inert gas concentration and oxide concentration. Experimental results for the heat flux and rate of temperature rise appear to be the most inconsistent

Non-contact Measurement of Remaining Thickness of Corroding Superheater Tubes. Phase 2

Energy Technology Data Exchange (ETDEWEB)

Borggreen, Kjeld; Storesund, Jan

2007-12-15

The Electro Magnetic Acoustic Transducer (EMAT) is a contactless thickness gauge for detection of corrosion on superheater tubes; it candidates as substitute for conventional manually operated contact UT transducers. It is the aim of the project to demonstrate the usefulness of two simple EMAT systems, Panametrics and Sonatest, for fast and reliable tube thickness inspections in difficult-to-access superheater sections. The present Phase 2 of the project involves testing of the systems on real compact superheaters in remote operation with the help of a mechanical manipulator designed and built for the purpose. The results are the following: - Both EMAT systems work well when tested in the field during handheld operation on tubes with a moderate thick layer of corrosion products and ash. The practical obtainable speed of cross-panel inspection of easily accessible panels is approximately 6 tubes per minute (6 thickness readings per minute). - The Sonatest system works well when tested in the field during remote operation on heavily corroded superheater tubes with thick ash layer. The Panametrics system was not found suitable for this type of field work. - The mechanical manipulator works well for cross-panel inspection of difficult-to-access superheater sections independent of the tube dimensions and the free space between the panels. In its present design it needs few improvements. - The practical obtainable speed of cross-panel inspection is 3 tubes per minute (3 thickness readings per minute). This speed is limited by the detection rate of the EMAT system and not by the travelling speed of the probe. - Scanning of tubes along their axis was never attempted, because the EMAT instruments were not capable of collecting data coming as a continuous stream. - It cannot be judged from visual alone and hardly from the service data, if a tube or a panel can be inspected by the magnetostrictive EMAT method or not. - The main contribution to failure of the EMAT inspection

SEM Investigation of Superheater Deposits from Biomass-Fired Boilers

DEFF Research Database (Denmark)

Jensen, Peter Arendt; Frandsen, Flemming; Hansen, Jørn

2004-01-01

, mature superheater deposit samples were extracted from two straw-fired boilers, Masnedø and Ensted, with fuel inputs of 33 MWth and 100 MWth, respectively. SEM (scanning electron microscopy) images and EDX (energy dispersive X-ray) analyses were performed on the deposit samples. Different strategies...

Overheating failure of superheater suspension tubes of a captive thermal power plant boiler

International Nuclear Information System (INIS)

Bhattacharya, Sova; Amir, Q.M.; Kannan, C.; Mahapatra, S.B.

2000-01-01

Failure of boiler tubes is the foremost cause of forced boiler outages. One of the predominant failure mechanism of boiler tubes is the stress rupture failure in the form of either short term overheating or long term overheating which are normally encountered in superheater and reheater sections working in the creep range. The strength of the boiler tube depends on the stress level as well on the temperature of exposure in the creep range. An increase in either can reduce the time to rupture. Time at the exposure temperature is an important factor based on which the failures are categorised as either short term or long term. Though there is no established time duration criteria demarcating the short or long term stress rupture failures, depending on the various manifestations on the failed samples, one can categorise the failure. This paper addresses one such stress rupture failure in the superheater section of a captive thermal power plant of a refinery. Multiple failures on the suspension coil of a superheater section was investigated for the cause of failure. Laboratory investigation of the failed sample involved visual inspection, dimensional measurements, chemical analysis of internal deposits and microstructural study. On the basis of these, the failure was attributed to deposition of trisodium phosphate carried over by the feed water into the superheater section resulting in chokage and increase in local operating hoop stresses of the tube. The ultimate failure was thus categorised as long term overheating failure. (author)

Nonlinear superheat and capacity control of a refrigeration plant

DEFF Research Database (Denmark)

Rasmussen, Henrik; Larsen, Lars F. S.

2009-01-01

This paper proposes a novel method for superheat and capacity control of refrigeration systems. A new low order nonlinear model of the evaporator is developed and used in a backstepping design of a nonlinear controller. The stability of the proposed method is validated theoretically by Lyapunov...

Prevention of superheater corrosion caused by chlorine; Tulistimien kloorikorroosion estaeminen

Energy Technology Data Exchange (ETDEWEB)

Roppo, J. [Kvaerner Power Oy, Tampere (Finland)

2006-12-19

Combustion of CO{sub 2}-neutral fuels is becoming more attractive and common method to decrease CO2 emissions of energy production. Also well managed and controlled combustion of waste fractions compared to their landfilling produces much less greenhouse gas emissions. In combustion of these fuels in high efficiency power plants notably increased superheater corrosion risk is prevailing, mainly caused by chlorine. Typical such fuels are forest, agricultural and household residues, biological sludge's of pulp and paper industry and RDF made from separated municipal and industrial solid waste. The goal of the project is to develop clearly cheaper and more effective method to protect superheaters, which enables combustion of biomass and waste fuels with higher energy shares. Tests in pilot and full scale power plants will reveal the potential and applicability of the developed method for commercial use. (orig.)

Research on the Superheater Material Properties for USC Boiler with 700°C Steam Parameter

Science.gov (United States)

Chongbin, Wang; Xueyuan, Xu; Yufeng, Zhu; Yongqiang, Jin; Hui, Tong; Yu, Wang; Xiaoli, Lu

This paper discusses the materials' properties of superheater for 700°C USC boiler, including Sanicro25, HR6W, 617mod and 740H, and analyzes the range of applicable temperature of superheater made of different tubes, such as T91, T92, Super304H, TP310HCbN, Sanicro25, HR6W, 617Mod and 740H. In addition, some suggestions on the material selection have been proposed.

Integrated boiler, superheater, and decomposer for sulfuric acid decomposition

Science.gov (United States)

Moore, Robert [Edgewood, NM; Pickard, Paul S [Albuquerque, NM; Parma, Jr., Edward J.; Vernon, Milton E [Albuquerque, NM; Gelbard, Fred [Albuquerque, NM; Lenard, Roger X [Edgewood, NM

2010-01-12

A method and apparatus, constructed of ceramics and other corrosion resistant materials, for decomposing sulfuric acid into sulfur dioxide, oxygen and water using an integrated boiler, superheater, and decomposer unit comprising a bayonet-type, dual-tube, counter-flow heat exchanger with a catalytic insert and a central baffle to increase recuperation efficiency.

Superheater corrosion in biomass-fired power plants: Investigation of Welds

DEFF Research Database (Denmark)

Montgomery, Melanie; Carlsen, B; Biede, O

2002-01-01

-fired Masnedø combined heat and power (CHP) plant to investigate corrosion at temperatures higher than that of the actual plant. The highest steam temperature investigated was 570°C. Various alloys of 12-22% chromium content were welded into this test loop. Their corrosion rates were similar and increased...... condense on superheater components. This gives rise to specific corrosion problems not previously encountered in coal-fired power plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A test superheater was built into the straw...... with temperature. The mechanism of attack was grain boundary attack as a precursor to selective chromium depletion of the alloy. In addition welds coupling various tubes sections were also investigated. It was seen that there was preferential attack around those welds that had a high nickel content. The welds...

KANDUNGAN INFORMASI PENGUMUMAN SAHAM BONUS: STUDI EMPIRIS DI BURSA EFEK INDONESIA

Directory of Open Access Journals (Sweden)

Lydia Angela Natasya

2017-03-01

Full Text Available The aim of this paper was to investigate the market reaction of bonus share announcement in Indonesia StockExchange. Bonus share was the signal given by company to public or stockholders. If bonus shares announcementconsisted of the information, it would be reacted by the market. There were pro’s and con’s about thefinding of bonus share’s announcement. The Standard of event study method had been used for the purpose ofstudying the bonus share issues announcement reaction. The proxies of market reaction were abnormal returnand trading volume activity. In this study, the researcher found a significant negative abnormal return and itmeant that the announcement of bonus share had negative information content. This finding probably meantthat most companies had liquidity problem. The study also found that the average of trading volume activitywas insignificantly decreased after bonus share announcement. This empirical study showed that bonus shareindicated a bad signal for the Indonesia market.

A fault tolerant superheat control strategy for supermarket refrigeration systems

DEFF Research Database (Denmark)

Vinther, Kasper; Izadi-Zamanabadi, Roozbeh; Rasmussen, Henrik

2013-01-01

, based on a maximum slope-seeking control method and only a single temperature sensor, is developed to drive the evaporator outlet temperature to a level that gives a suitable superheat of the refrigerant. The FTC strategy requires no a priori system knowledge or additional hardware and functions...

Tuning and performance evaluation of PID controller for superheater steam temperature control of 200 MW boiler using gain phase assignment algorithm

Science.gov (United States)

Begum, A. Yasmine; Gireesh, N.

2018-04-01

In superheater, steam temperature is controlled in a cascade control loop. The cascade control loop consists of PI and PID controllers. To improve the superheater steam temperature control the controller's gains in a cascade control loop has to be tuned efficiently. The mathematical model of the superheater is derived by sets of nonlinear partial differential equations. The tuning methods taken for study here are designed for delay plus first order transfer function model. Hence from the dynamical model of the superheater, a FOPTD model is derived using frequency response method. Then by using Chien-Hrones-Reswick Tuning Algorithm and Gain-Phase Assignment Algorithm optimum controller gains has been found out based on the least value of integral time weighted absolute error.

The Phenomenon of Superheat of Liquids: In Memory of Vladimir P. Skripov

Science.gov (United States)

Skripov, P. V.; Skripov, A. P.

2010-05-01

This article is devoted to the memory of Vladimir P. Skripov (1927-2006). He has received worldwide recognition for his monograph on metastable liquids published in 1972 (the English edition was published in 1974). We briefly discuss some studies deal with the phenomenon of attainable superheat of liquids and with measurements of thermophysical properties of liquids under conditions of a moderate degree of superheat. Main attention is paid to the studies performed by V.P. Skripov and his research group in the 1960s and 1970s. Experimental methods which provided break-throughs in research on both spontaneous boiling-up kinetics and substance properties (the specific volume, isobaric heat capacity, ultrasound speed, and viscosity) in super-heated states are presented.

SEBARAN STASIONER PADA SISTEM BONUS-MALUS SWISS SERTA MODIFIKASINYA (Stationary Distribution of Swiss Bonus-Malus System and its Modification

Directory of Open Access Journals (Sweden)

Cherry Galatia Ballangan

2002-01-01

Full Text Available Bonus-Malus System is a system in actuary that introduce the premium class (state partition, where the state is influenced by the number of annual claims reported by the policy holder. We could base the determination of the state on the stationary distribution that represent the number of policy holders in any state. Swiss Bonus-Malus System has 22 state. The number of state that involved in this system result in the difficulty of stationary distribution determination. Therefore, the aim of this paper is to study a method to obtain stationary distribution of Swiss Bonus-Malus System by recursive formula, with this recursive formula, the stationary distribution of Swiss Bonus-Malus System can be determined easier. Modification of this system with infinite state result in the changes of recursive formula to obtain the stationary. This changes including the determining of base value of the recursive formula. Abstract in Bahasa Indonesia : Sistem Bonus-Malus merupakan sistem dalam aktuaria yang memperkenalkan pembagian kelas premi (state yang dipengaruhi oleh jumlah klaim yang diajukan oleh pemegang polis tiap tahunnya. Penetapan state dalam sistem ini didasarkan pada pencarian sebaran stasioner yang menyatakan banyaknya pemegang polis dalam tiap state. Sistem Bonus-Malus Swiss (BMS memiliki 22 state. Banyaknya state yang terlibat dalam sistem ini mengakibatkan sulitnya penentuan sebaran stasioner pada sistem BMS tersebut. Karena itu dalam tulisan ini dipelajari suatu metode penentuan sebaran stasioner dari sistem BMS tersebut, yaitu dengan menggunakan formula rekursif. Dengan formula rekursif ini, sebaran stasioner sistem BMS dapat ditentukan dengan mudah. Modifikasi sistem BMS untuk jumlah state yang tak hingga mengakibatkan perubahan pada formula rekursif untuk mencari sebaran stasionernya. Perubahan ini meliputi penetapan nilai awal dari formula rekursif tersebut. Kata kunci: sebaran stasioner, formula rekursif, sistem Monus-Malus Swiss.

A Stochastic mesoscopic model for predicting the globular grain structure and solute redistribution in cast alloys at low superheat

International Nuclear Information System (INIS)

Nastac, Laurentiu; El Kaddah, Nagy

2012-01-01

It is well known that casting at low superheat has a strong influence on the solidification morphology and macro- and microstructures of the cast alloy. This paper describes a stochastic mesoscopic solidification model for predicting the grain structure and segregation in cast alloy at low superheat. This model was applied to predict the globular solidification morphology and size as well as solute redistribution of Al in cast Mg AZ31B alloy at superheat of 5°C produced by the Magnetic Suspension Melting (MSM) process, which is an integrated containerless induction melting and casting process. The castings produced at this low superheat have fine globular grain structure, with an average grain size of 80 μm, which is about 3 times smaller than that obtained by conventional casting techniques. The stochastic model was found to reasonably predict the observed grain structure and Al microsegregation. This makes the model a useful tool for controlling the structure of cast magnesium alloys.

Preventing superheater corrosion by additives; Tulistimien kloorikorroosion estaeminen lisaeainein

Energy Technology Data Exchange (ETDEWEB)

Aho, M.; Vainikka, P. [VTT, Espoo (Finland); Skrifvars, B.J.; Yrjas, P. [Aabo Akademi, Process Chemistry, Turku (Finland)

2006-12-19

The new superheater protection methods enable combustion of demanding biomass with higher portions than at present. This benefit reduces CO{sub 2} emissions from energy production and the use of demanding biomass in energy production will extend replacing biowaste landfilling with strong CH{sub 4} formation. The results assist also to meet the goals of the use of logging residues in energy production in Finland. (orig.)

CLUMPED LIGHT WATER MODERATED UO$sub 2$ SUPERHEAT CRITICALS. PART II. ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

Petersen, G. T.

1963-11-15

Critical and subcritical reactivity measurements on an EVESR-type core, using EVESR UO/sub 2/ superheat fuel elements, are analyzed in order to obtain a physics design model for use in the EVESR. (T.F.H.)

Some post operational adjustments to the prototype fast reactor at Dounreay

International Nuclear Information System (INIS)

Lunt, A.R.W.

1979-01-01

Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

Massachusetts Signing Bonus Program for New Teachers

Directory of Open Access Journals (Sweden)

R. Clarke Fowler

2003-04-01

Full Text Available This article examines the Massachusetts Signing Bonus Program for New Teachers, a nationally prominent program that has recruited and prepared $20,000 bonus recipients to teach after seven weeks' training at the Massachusetts Institute for New Teachers (MINT. Although state officials have trumpeted this initiative as a national model that other states are copying, they announced in November 2002 that they were radically changing it. The changes included halting the state's national recruitment efforts and replacing the seven-week, fast-track training program designed by the New Teacher Project with year-long programs to be designed by three of the state's education schools. Even though the state spent more than $50,000 recruiting individuals from states outside the Northeast over the first four program years, it garnered just seven bonus recipients from the non-Northeast states its recruiters visited, only four of whom were still teaching in Fall 2002. The state did, however, generate a substantial number of applicants in each program year (ranging from 783 to nearly 950, most of whom came from Massachusetts or nearby states. Contrary to state officials' claims, though, it appears that many of these individuals had substantial prior educational experience. Although officials stated that all bonus teachers would go to 13 designated high-need urban districts, the state has never met this commitment, sending fewer bonus teachers to these districts in each of the first three years of the program. The state has lost a high percentage of its bonus teachers to attrition particularly in state-designated, high-need districts. These attrition rates are substantially higher than comparable national rates. Although the state has portrayed the Bonus and MINT programs, combined, as highly successful, officials exaggerated many of the purported positive outcomes. On the positive side, independent survey data (Churchill et al., 2002 indicated that principals

Don't forget the gravy! : are bonuses and time rates complements?

OpenAIRE

Green, Colin; Heywood, John

2016-01-01

The press often depicts bonuses as extra payments to the already well compensated and calls for reform. Yet, these calls typically ignore the efficiency argument that bonuses are potentially risky performance pay that substitute for salary compensation. This paper uses representative UK data to estimate that bonuses appear not to substitute for salary in cross-sectional estimates. Yet, when controlling for time invariant characteristics in panel data, bonuses emerge as partial substitutes. Ea...

Teacher Salary Bonuses in North Carolina. Research Brief

Science.gov (United States)

National Center on Performance Incentives, 2008

2008-01-01

In "Teacher Salary Bonuses in North Carolina"--a paper presented at the February 2008 National Center on Performance Incentives research to policy conference--Jacob Vigdor of Duke University reviews a teacher salary bonus program operating in North Carolina. Known officially as the ABC's of Public Education, the program awards teachers…

Superheater fouling in a BFB boiler firing wood-based fuel blends

NARCIS (Netherlands)

Stam, A.F.; Haasnoot, K.; Brem, Gerrit

2014-01-01

Four different fuel blends have been fired in a 28 MWel BFB. Wood pellets (test 0) were not problematic for about ten years, contrary to a mixture of demolition wood, wood cuttings, compost overflow, paper sludge and roadside grass (test 1) which caused excessive fouling at a superheater bundle

FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

Energy Technology Data Exchange (ETDEWEB)

Hwang, Gisuk [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States); Kaviany, Massoud [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Moriyama, Kiyofumi [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hwang, Byoungcheol; Lee, Mooneon; Kim, Eunho; Park, Jin Ho [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Nasersharifi, Yahya [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States)

2016-08-15

Highlights: • The numerical approach for FARO experimental data is suggested. • The cooling mechanism of ex-vessel corium is suggested. • The predicted minimum pool depth for no cake formation is suggested. - Abstract: The FARO tests have aimed at understanding an important severe accident mitigation action in a light water reactor when the accident progresses from the reactor pressure vessel boundary. These tests have aimed to measure the coolability of a molten core material (corium) gravity dispersed as jet into a water pool, quantifying the loose particle diameter distribution and fraction converted to cake under range of initial melt superheat and pool temperature and depth. Under complete hydrodynamic breakup of corium and consequent sedimentation in the pool, the initially superheated corium can result in debris bed consisting of discrete solid particles (loose debris) and/or a solid cake at the bottom of the pool. The success of the debris bed coolability requires cooling of the cake, and this is controlled by the large internal resistance. We postulate that the corium cake forms when there is a remelting part in the sediment. We show that even though a solid shell forms around the melt particles transiting in the water pool due to film-boiling heat transfer, the superheated melt allows remelting of the large particles in the sediment (depending on the water temperature and the transit time) using the COOLAP (Coolability Analysis with Parametric fuel-cooant interaction models) code. With this remelting and its liquid-phase sintering of the non-remelted particles, we predict the fraction of the melt particles converting to a cake through liquid sintering. Our predictions are in good agreement with the existing results of the FARO experiments. We address only those experiments with pool depths sufficient/exceeding the length required for complete breakup of the molten jet. Our analysis of the fate of molten corium aimed at devising the effective

Steam generator materials constraints in UK design gas-cooled reactors

International Nuclear Information System (INIS)

James, D.W.

1988-01-01

A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

Influence of surface roughness and melt superheat on HDA process to form a tritium permeation barrier on RAFM steel

Energy Technology Data Exchange (ETDEWEB)

Purushothaman, J. [B.S. Abdur Rahman University, Chennai 600048 (India); MTD, MMG, IGCAR, Kalpakkam 603102 (India); Ramaseshan, R., E-mail: seshan@igcar.gov.in [TFCS, SND, MSG, IGCAR, Kalpakkam 603102 (India); Albert, S.K. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Rajendran, R. [B.S. Abdur Rahman University, Chennai 600048 (India); Gowrishankar, N. [IP Rings Ltd., Maraimalainagar, Chennai 603209 (India); Ramasubbu, V. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Murugesan, S.; Dasgupta, Arup [PMG, MMG, IGCAR, Kalpakkam 603102 (India); Jayakumar, T. [MTD, MMG, IGCAR, Kalpakkam 603102 (India)

2015-12-15

Highlights: • Surface modified RAFMS samples were subjected to HDA and thermal oxidation. • Sample modified by SB process showed better coating and interface morphology. • Aluminized samples at 740 °C for 2 min showed Fe{sub 2}Al{sub 9}Si{sub 2} intermetallic phase. • Oxidized samples showed Fe{sub 2}Al{sub 8}Si, Fe{sub 2}Al{sub 3}Si{sub 3} and Fe{sub 3}Al{sub 2}Si{sub 3} intermetallic phases. • A uniform permeation barrier Al{sub 2}O{sub 3} was formed on the coating of oxidized HDA samples. - Abstract: The most optimal candidate material for fabrication of Test Blanket Module (TBM) in the installation of ITER and future fusion reactors is Reduced Activation Ferritic Martensitic (RAFM) steel, yet one of the major challenges that need to be addressed with RAFM is minimizing the loss of tritium in a reactor environment through the formation of tritium permeation barrier. One of the most promising methods for the tritium permeation barrier is through duplex coating with Al{sub 2}O{sub 3}/Fe–Al which is well known to reduce tritium permeation rate by several orders of magnitude. The present work aims to form an alumina layer on RAFM steel by a two-step method, which consists of (i) Hot Dip Aluminizing (HDA) and (ii) conversion of Al into alumina by a subsequent oxidation process. In addition, the influence of surface roughness of the substrate, superheat condition of the Al alloy melt and its composition on microstructural properties of coating before and after oxidation were investigated using OM, SEM–EDS, XRD, indentation micro hardness and scratch test. The experimental results confirmed the formation of alumina layer on RAFM steel after the HDA and oxidation process. Moreover, the surface roughness of the substrate, melt superheat of Al alloy and its composition are found to have a significant influence on the microstructure, thickness, micro-hardness, nature of intermetallic compounds formed and adhesion strength of the coating.

Dynamic performances of wet turbine and steam-separator-superheater and their mathematical simulation as objects of temperature control

International Nuclear Information System (INIS)

Golovach, E.A.

1985-01-01

A mathematical model of a turbine and steam-separator-superheater (SSS) as applied to solution of the tasks of steam temperature regulaton after SSS has been developed. SSS as objects of steam temperature control are considerably less inertial, than intermediate superheaters (IS) of power units in thermal power plants, since for typical SSS and IS considered the duration of transition process according to steam temperature after SSS is 5-10 times loweA than for IS

Non-contact Measurement of Remaining Thickness of Corroding Superheater Tubes. Phase 1

Energy Technology Data Exchange (ETDEWEB)

Borggreen, Kjeld; Storesund, Jan

2006-10-15

Corrosion of superheaters has become a severe problem in many biomass boilers and incineration plants. This new situation calls for frequent tube wall thickness testing of the superheaters during very short shut-downs. To meet this demand Electro Magnetic Acoustic Transducer (EMAT) candidates as substitute for conventional manually operated contact UT-transducers. The EMAT can contactlessly generate an ultrasonic wave in the interphase between the external oxide and the metal. This means that measurements can be undertaken much quicker and with a much higher coverage simultaneously, without preceding blast operations. It is the aim of the project to demonstrate the usefulness of two simple EMAT systems, Panametrics and Sonatest, for fast and reliable tube thickness inspections in difficult-to-access superheater sections. The present Phase 1 of the project involves testing of the performance of the two systems in laboratory with the following results: 1. Both instruments work well on plate, tube, and pipe samples assuming the presence of an external oxide layer formed at a temperature above approximately 400 deg C. 2. Both instruments work well on all types of ferritic and martensitic steels but not on austenitic steels. 3. Both instruments work well independent of the thickness of the active oxide layer. 4. Both instruments work well independent of tube diameter, wall thickness, and sample width. 5. Both instruments work well over a very large range of wall thicknesses. Minimum tube wall thickness is less than 1.8 mm. 6. The tolerable lift-off (free distance between transducer and tube surface) is 2.4 - 3.0 mm for Panametrics system and 3.6 - 4.8 mm for Sonatest's system. The tolerable lift-off is a measure of the thickness of ash deposits, which can be tolerated on the tube surface as well as the misplacement, which can be tolerated in case of remote tube testing. 7. The tolerable off-set between tube axis and probe axis is very large for both instruments (10

Soft Sensor for Oxide Scales on the Steam Side of Superheater Tubes under Uneven Circumferential Load

Directory of Open Access Journals (Sweden)

Qing Wei Li

2015-01-01

Full Text Available A soft sensor for oxide scales on the steam side of superheater tubes of utility boiler under uneven circumferential loading is proposed for the first time. First finite volume method is employed to simulate oxide scales growth temperature on the steam side of superheater tube. Then appropriate time and spatial intervals are selected to calculate oxide scales thickness along the circumferential direction. On the basis of the oxide scale thickness, the stress of oxide scales is calculated by the finite element method. At last, the oxide scale thickness and stress sensors are established on support vector machine (SMV optimized by particle swarm optimization (PSO with time and circumferential angles as inputs and oxide scale thickness and stress as outputs. Temperature and stress calculation methods are validated by the operation data and experimental data, respectively. The soft sensor is applied to the superheater tubes of some power plant. Results show that the soft sensor can give enough accurate results for oxide scale thickness and stress in reasonable time. The forecasting model provides a convenient way for the research of the oxide scale failure.

A Statistical Assessment of Demographic Bonus towards Poverty Alleviation

Directory of Open Access Journals (Sweden)

Jamal Abdul Nasir

2011-09-01

Full Text Available The shift of birth and death rates from high to low level in any population is referred as demographic transition. Mechanically, the transition of a society creates more working member of its own population commonly called demographic bonus. This articleempirically explores the realistic soundness of demographic bonus in reducing the poverty level of the society. Three contrasting regions namely Eastern Asia, Central America and Oceania were selected for analytical purposes. The findings indicate that Eastern Asia and Oceania are currently facing the end of their transition whereas theCentral America is lagged behind in transition. Central America due to last runner in transition race is the sustained recipient of its own demographic bonus by the year 2030.On the basis of three mechanisms namely: labour supply, savings and human capital, the Eastern Asian region is found to be successful beneficiary of its own demographic gift which concludes that many million people have escaped from poverty. Under the right policy environment on the above three mechanisms, Eastern Asia experience indicates the realistic contribution of demographic bonus to reduce poverty.

A steam superheater exchanger provided with two coaxial casings and an horizontal axis

International Nuclear Information System (INIS)

Marjollet, Jacques; Palacio, Gerard; Tondeur, Gerard.

1976-01-01

This invention concerns the general lay-out of an horizontal axis separator-superheater for supplying steam to a high power turbine, particularly for a nuclear power station. The invention significantly reduces the length of the pipework connecting the superheated steam outlet and its inlet to the turbine. For this, the outer casing is provided with a coaxial internal annular sleeve in which are housed, one above the other, the separator and the bundle of superheater tubes through which circulates the water emulsion to be separated and steam to be superheated. At the end of its treatment, the superheated steam spreads out in the space between the sleeve and the outer casing from whence it can be drawn off at any point of its periphery, thus making it possible to choose an extraction point as near as possible to the inlet of the turbine to be fed [fr

Reflection of Loyalty Programmes with the Use of Bonuses in Accounting

OpenAIRE

Sakharov Pavlo O.

2013-01-01

The article studies loyalty programmes that are used at modern trading enterprises and discovers main barriers on the way of development of methodical approaches of bonus accounting provided to the clients of loyalty programmes. It analyses theoretical and methodical aspects of bonus accounting in international practice and methods of reflection of loyalty programmes in accounting with the use of bonuses existing in Ukraine. Comparing electronic money and settlement elements of the loyalty pr...

The Bonus Army: A Lesson on the Great Depression

Science.gov (United States)

Chiodo, John J.

2011-01-01

After the end of World War I, Congress enacted a bill that would reward military veterans for their service. The bill provided the veterans cash bonuses that would be paid starting in 1945. But as the nation settled into the Great Depression these veterans began to clamor for payment of their bonuses. In May of 1932, and estimated 15,000 veterans…

Bonus payments as an anti-corruption instrument: A theoretical approach

KAUST Repository

Cracau, Daniel; Franz, Benjamin

2013-01-01

We analyse bonus payments for officials, who transfer payments truthfully to the government rather than collecting bribes. We show that optimised bonus payments are always beneficial to the government, making them a more effective anti-corruption measure than simple wage increases. © 2013 Elsevier B.V.

Bonus payments as an anti-corruption instrument: A theoretical approach

KAUST Repository

Cracau, Daniel

2013-07-01

We analyse bonus payments for officials, who transfer payments truthfully to the government rather than collecting bribes. We show that optimised bonus payments are always beneficial to the government, making them a more effective anti-corruption measure than simple wage increases. © 2013 Elsevier B.V.

Find Shortage Areas: HPSAs Eligible for the Medicare Physician Bonus Payment

Data.gov (United States)

U.S. Department of Health & Human Services — The HPSAs Eligible for the Medicare Physician Bonus Payment advisor tools allows the user (physician) to determine if an address is eligible for bonus payments....

Probabilistic approach to determining the optimum replacement of a superheater stage in 680 MW coal-fired boiler

Energy Technology Data Exchange (ETDEWEB)

Bos, Robert; Star, Ruud van der [Nuon Power Generation, Amsterdam (Netherlands)

2009-07-01

The boiler of the NUON power plant HW08 that went into operation in 1993 is designed as Benson boiler and mainly fired with hard coal. A creep-related tube failure occurred in the tertiary superheater that had been due to increased wall temperature caused by steam side formation of oxide layers. The theoretical lifetime of the components was calculated with the aid of the results of steam side oxide measurements and condition evaluation of the tertiary superheater with the aid of tube samples. The objective is to establish an operation and maintenance schedule for the desired operating lifetime of 300,000 hours. (orig.)

Improved superheater tubing material - Ti and Nb bearing austenitic steel

International Nuclear Information System (INIS)

Kinoshita, K.; Mimino, T.; Minegishi, I.

1975-01-01

A newly developed 18 Cr-8 Ni stainless steel modified with small amounts of Ti and Nb has considerably high stress-rupture strength and is considered to be suitable for superheater material for power boilers. Data for stress-rupture and creep for long times, the strength of welded joints, the changes of characteristics due to exposure to high temperatures, etc., are presented and discussed. Some investigations after trial services indicate that the experimental data are applicable to actual applications. (author)

29 CFR 778.212 - Gifts, Christmas and special occasion bonuses.

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Gifts, Christmas and special occasion bonuses. 778.212... COMPENSATION Payments That May Be Excluded From the âRegular Rateâ Bonuses § 778.212 Gifts, Christmas and... “regular rate” shall not be deemed to include “sums paid as gifts; payments in the nature of gifts made at...

Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

International Nuclear Information System (INIS)

Smith, J.; Hart, R.; Lazic, L.

2009-01-01

Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

Improving the Navy's Officer Bonus Program Effectiveness

National Research Council Canada - National Science Library

Filip, William N

2006-01-01

.... Combining the information gained from current policies and research already conducted in the academic arena, the author proposes a workable bonus structure to meet the recruitment and retention goals...

41 CFR 101-25.103-3 - Trading stamps or bonus goods received from contractors.

Science.gov (United States)

2010-07-01

... 41 Public Contracts and Property Management 2 2010-07-01 2010-07-01 true Trading stamps or bonus... PROCUREMENT 25-GENERAL 25.1-General Policies § 101-25.103-3 Trading stamps or bonus goods received from contractors. When contracts contain a price reduction clause, any method (such as trading stamps or bonus...

Evaluation of property tax bonus to promote solar thermal systems in Andalusia (Spain)

International Nuclear Information System (INIS)

Sánchez-Braza, Antonio; Pablo-Romero, María del P.

2014-01-01

This paper evaluates the effects of a property tax bonus to promote the installation of solar–thermal energy systems in buildings in Andalusia (southern Spain). The propensity score matching methodology is used. The treatment group consists of municipalities of Andalusia that established property tax bonuses in their municipalities in 2010. The control group consists of municipalities that did not. The response variable measures the number of new square meters of solar thermal systems installed in 2010. The analysis leads to the conclusion that municipalities that established a property tax bonus had installed, on average, 102.245 to 122.389 square meters more. These results indicate that the percentage increase in squares meters installed in municipalities which adopted the tax bonus promotion ranged from 70.74% to 98.38%. These percentages were lower for rural municipalities (49.00% to 77.06%). - Highlights: • This paper evaluates the effects of a tax bonus to promote solar–thermal energy. • We analyse the effect of this measure for 585 Andalusia municipalities. • The propensity score-matching methodology is used. • The percentage increase of square meters installed ranged from 70.74% to 98.38%. • Tax bonus was an effective tool to promote solar thermal in Andalusia

Thermal load non-uniformity estimation for superheater tube bundle damage evaluation

Directory of Open Access Journals (Sweden)

Naď Martin

2018-01-01

Full Text Available Industrial boiler damage is a common phenomenon encountered in boiler operation which usually lasts several decades. Since boiler shutdown may be required because of localized failures, it is crucial to predict the most vulnerable parts. If damage occurs, it is necessary to perform root cause analysis and devise corrective measures (repairs, design modifications, etc.. Boiler tube bundles, such as those in superheaters, preheaters and reheaters, are the most exposed and often the most damaged boiler parts. Both short-term and long-term overheating are common causes of tube failures. In these cases, the design temperatures are exceeded, which often results in decrease of remaining creep life. Advanced models for damage evaluation require temperature history, which is available only in rare cases when it has been measured and recorded for the whole service life. However, in most cases it is necessary to estimate the temperature history from available operation history data (inlet and outlet pressures and temperatures etc.. The task may be very challenging because of the combination of complex flow behaviour in the flue gas domain and heat transfer phenomena. This paper focuses on estimating thermal load non-uniformity on superheater tubes via Computational Fluid Dynamics (CFD simulation of flue gas flow including heat transfer within the domain consisting of a furnace and a part of the first stage of the boiler.

KLM senior managers forgo controversial bonuses

NARCIS (Netherlands)

Grünell, M.

2004-01-01

In April 2004, the members of the management board of KLM waived bonuses offered to them in the run-up to the Dutch airline's merger with Air France, under pressure from trade unions and the government. During the current agreed national wage freeze, increases in senior management remuneration are

Evaporator Superheat Control With One Temperature Sensor Using Qualitative System Knowledge

DEFF Research Database (Denmark)

Vinther, Kasper; Hillerup Lyhne, Casper; Baasch Sørensen, Erik

2012-01-01

This paper proposes a novel method for superheat control using only a single temperature sensor at the outlet of the evaporator, while eliminating the need for a pressure sensor. An inner loop controls the outlet temperature and an outer control loop provides a reference set point, which is based...... filling of the evaporator, with only one temperature sensor. No a priori model knowledge was used and it is anticipated that the method is applicable on a wide variety of refrigeration systems....

Financial gains and risks in pay-for-performance bonus algorithms.

Science.gov (United States)

Cromwell, Jerry; Drozd, Edward M; Smith, Kevin; Trisolini, Michael

2007-01-01

Considerable attention has been given to evidence-based process indicators associated with quality of care, while much less attention has been given to the structure and key parameters of the various pay-for-performance (P4P) bonus and penalty arrangements using such measures. In this article we develop a general model of quality payment arrangements and discuss the advantages and disadvantages of the key parameters. We then conduct simulation analyses of four general P4P payment algorithms by varying seven parameters, including indicator weights, indicator intercorrelation, degree of uncertainty regarding intervention effectiveness, and initial baseline rates. Bonuses averaged over several indicators appear insensitive to weighting, correlation, and the number of indicators. The bonuses are sensitive to disease manager perceptions of intervention effectiveness, facing challenging targets, and the use of actual-to-target quality levels versus rates of improvement over baseline.

Modeling and performance of Bonus-Malus Systems: Stationarity versus age-correction

DEFF Research Database (Denmark)

Asmussen, Søren

In a bonus-malus system in car insurance, the bonus class of a customer is updated from a year to the next as a function of the current class and the number of claims in the year (assumed Poisson). Thus the sequence of classes of a customer in consecutive years forms a Markov chain, and most of t...

The effect of bonus/malus systems on the sales of new vehicles; Wirkung von Bonus-Malus-Systemen beim Neuwagenkauf

Energy Technology Data Exchange (ETDEWEB)

Haan, P. de

2008-07-01

This presentation made at the Swiss 2008 research conference on traffic by Peter de Haan from the Institute for Environmental Decisions at the Swiss Federal Institute of Technology ETH in Zurich takes a look at the effects of bonus/malus systems on the sales of new vehicles. The author notes that in spite of technological progress made, increasing car size, weight and mileage along with increased motor power has eaten up two-thirds of energy-efficiency progress made. The effect of the long lifetime of modern vehicles is commented on. The results of a survey made are presented and discussed, as are the results of simulations. Further aspects discussed include effects on the motor trade and road safety and the indirect effects of bonuses. Finally, trends in the area are noted.

On-line Auto-Tuning of PI Control of the Superheat for a Supermarket Refrigeration System

DEFF Research Database (Denmark)

Yang, Zhenyu; Andersen, Casper; Izadi-Zamanabadi, Roozbeh

2011-01-01

An online PI auto-tuning method is proposed for superheat control for a type of supermarket refrigeration systems. The proposed procedure consists of three serial steps: Step-One uses one of the two proposed empirical methods, namely multi-step method and relay method, for modeling initialization...

29 CFR 778.211 - Discretionary bonuses.

Science.gov (United States)

2010-07-01

... without prior promise or agreement. The employee has no contract right, express or implied, to any amount... of the period and not pursuant to any prior contract, agreement, or promise causing the employee to... hand, if a bonus such as the one just described were paid without prior contract, promise or...

Can the lifetime of the superheater tubes be predicted according to the fuel analyses? Assessment from field and laboratory data

Energy Technology Data Exchange (ETDEWEB)

Salmenoja, K [Kvaerner Pulping Oy, Tampere (Finland)

1999-12-31

Lifetime of the superheaters in different power boilers is more or less still a mystery. This is especially true in firing biomass based fuels (biofuels), such as bark, forest residues, and straw. Due to the unhomogeneous nature of the biofuels, the lifetime of the superheaters may vary from case to case. Sometimes the lifetime is significantly shorter than originally expected, sometimes no corrosion even in the hottest tubes is observed. This is one of the main reasons why the boiler operators often demand for a better predictability on the corrosion resistance of the materials to avoid unscheduled shutdowns. (orig.) 9 refs.

Can the lifetime of the superheater tubes be predicted according to the fuel analyses? Assessment from field and laboratory data

Energy Technology Data Exchange (ETDEWEB)

Salmenoja, K. [Kvaerner Pulping Oy, Tampere (Finland)

1998-12-31

Lifetime of the superheaters in different power boilers is more or less still a mystery. This is especially true in firing biomass based fuels (biofuels), such as bark, forest residues, and straw. Due to the unhomogeneous nature of the biofuels, the lifetime of the superheaters may vary from case to case. Sometimes the lifetime is significantly shorter than originally expected, sometimes no corrosion even in the hottest tubes is observed. This is one of the main reasons why the boiler operators often demand for a better predictability on the corrosion resistance of the materials to avoid unscheduled shutdowns. (orig.) 9 refs.

Complementary Methods for the Characterization of Corrosion Products on a Plant-Exposed Superheater Tube

DEFF Research Database (Denmark)

Okoro, Sunday Chukwudi; Nießen, Frank; Villa, Matteo

2017-01-01

In this work, complex corrosion products on a superheater tube exposed to biomass firing were characterized by the complementary use of energy-dispersive synchrotron diffraction, electron microscopy, and energy-dispersive X-ray spectroscopy. Non-destructive synchrotron diffraction in transmission......-rich austenite phase to selective removal of Fe and Cr from the alloy, via a KCl-induced corrosion mechanism. Compositional variations were related to diffraction results and revealed a qualitative influence of the spinel cation concentration on the observed diffraction lines.......In this work, complex corrosion products on a superheater tube exposed to biomass firing were characterized by the complementary use of energy-dispersive synchrotron diffraction, electron microscopy, and energy-dispersive X-ray spectroscopy. Non-destructive synchrotron diffraction in transmission...... geometry measuring with a small gauge volume from the sample surface through the corrosion product allowed depth-resolved phase identification and revealed the presence of (Fe,Cr)2O3 and FeCr2O4. This was supplemented by microstructural and elemental analysis correlating the additional presence of a Ni...

Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

OpenAIRE

Al Hajri, Mohammed; Malik, Anees U.; Meroufel, Abdelkader; Al-Muaili, Fahd

2015-01-01

Dissimilar metal weld (DMW) joint between alloyed steel (AS) and stainless steel (SS) failed at one of intermediate temperature superheater (ITSH) tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years) where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results o...

Thermal-structural response of EBR-II major components under reactor operational transients

International Nuclear Information System (INIS)

Chang, L.K.; Lee, M.J.

1983-01-01

Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

The influence of year-end bonuses on colorectal cancer screening.

Science.gov (United States)

Armour, Brian S; Friedman, Carol; Pitts, M Melinda; Wike, Jennifer; Alley, Linda; Etchason, Jeff

2004-09-01

To estimate the effect of physician bonus eligibility on colorectal cancer (CRC) screening, controlling for patient and primary care physician characteristics. Retrospective study using managed care plan claims data from 2000 and 2001. Data on 50-year-old commercially insured patients in a managed care health plan were linked to enrollment and provider files. The data included information on 6749 patients (3058 in 2000 and 3691 in 2001). Multivariate logistic regression models were used to assess the association between CRC screening receipt and physician bonus eligibility. From 2000 to 2001, CRC screening use increased from 23.4% to 26.4% (P analysis revealed that the probability that a patient received a CRC screening was approximately 3 percentage points higher in the bonus year, 2001 (P < .01). Bonuses targeted at individual physicians were associated with increased use of CRC screening tests. However, more research is needed to examine the effect of performance-based incentives on resource use and the quality of medical care. Specifically, there is a need to determine whether explicit financial incentives are effective in reducing racial disparities in the quality of patient care. This has particular relevance for CRC screening given that black patients are less likely to be screened, they have higher CRC incidence and mortality rates compared with other racial groups, and screening has been shown to be more cost effective in this population.

26 CFR 1.612-3 - Depletion; treatment of bonus and advanced royalty.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 7 2010-04-01 2010-04-01 true Depletion; treatment of bonus and advanced... (CONTINUED) INCOME TAX (CONTINUED) INCOME TAXES (CONTINUED) Natural Resources § 1.612-3 Depletion; treatment... the payee as a cost depletion deduction in respect of the bonus an amount equal to that proportion of...

Estudio comparativo entre metodologías para el diseño de Sistemas Bonus-Malus.

Directory of Open Access Journals (Sweden)

Núñez del Prado, José Antonio

2004-01-01

Full Text Available El objetivo de este trabajo es mostrar un método para hacer comparaciones entre Sistemas Bonus Malus. Se comparan Sistemas Bonus Malus obtenidos mediante la metodología GPBM, basada en Programación por Metas y Sistemas Bonus Malus obtenidos mediante la metodología clásica, conocida como Escala de Bayes.

Bonus and Rebate: Perception of Exclusive GSM Dealers and Their Salesmen in Turkey

Directory of Open Access Journals (Sweden)

Melik Karabiyikoglu

2012-07-01

Full Text Available The aim of this paper is to study bonus and rebate for exclusive dealers in GSM sector in Turkey, for which operators and handset manufacturers provide special incentives to dealers for exceeding specific sales targets. They develop a theoretical model of dealers and manufacturer behavior based on observations about key aspects of the mobile phone market. The analysis provides important insights about sales bonus. For example, rebate is not preferred by the owners of dealers instead of bonuses. The authors find empirical support when they test the theoretical results.

Are CEOs incentivized to avoid Corporate Taxes? - Empirical Evidence on Managerial Bonus Contracts

NARCIS (Netherlands)

H. Schmittdiel (Heiner)

2014-01-01

markdownabstract__Abstract__ In this paper, we test empirically whether there is a relationship between corporate income taxes and CEO bonus payments. Using Compustat and ExecuComp data from 1992 to 2010, we find mixed results. Looking at the whole sample, the average bonus contract rewards

Growth rates of breeder reactor fuel. Final report

International Nuclear Information System (INIS)

Ott, K.O.

1979-01-01

During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus

Incoloy 800 stands up to radiation and corrosion in high temperature gas cooled reactors

International Nuclear Information System (INIS)

Anon.

1975-01-01

Incoloy 800 has been selected for heat exchangers in helium cooled nuclear reactor prototypes for exposure to 350 to 800 0 C helium and high temperature high purity water and steam. 304H stainless steel used in heat exchangers in original design cracked in the superheater area, bellows and tubing after static pressure tests but before exposure to steam. Residual stress, chlorides, and oxygen were deduced to have caused the failures

23 CFR 190.5 - Bonus project claims.

Science.gov (United States)

2010-04-01

... which advertising controls are in effect. The eligible system mileage reported in subsequent projects on... CONTROLLING OUTDOOR ADVERTISING ON THE INTERSTATE SYSTEM § 190.5 Bonus project claims. (a) The State may claim payment by submitting a form PR-20 voucher, supported by strip maps which identify advertising control...

The experimental and engineering programmes to support the PFR Safety Case following the Superheater 2 under sodium leak: In particular, large scale experiments in the Super Noah Rig at Dounreay

International Nuclear Information System (INIS)

Currie, R.; Henderson, J.D.C.

1990-01-01

The original safety Case for the Prototype Fast Reactor (PFR) at Dounreay was based on the Double-ended-guillotine failure (DEGF) of one tube followed by six more DEGFs spread out at 3s intervals. Because the DEGF flowrate in the Evaporator units was considerably greater than those for the Superheater and Reheater units, pressure loading predictions were based on a leak incident in the Evaporator. As data became available from sodium-water reaction experiments, this Design Basis Accident (DBA) was revised to be the failure of a single tube (1DEGF). Pressure loadings for the plant were still based on the Evaporator. The plant was, however, designed against the original DBA of 1+6 DEGFs. The under sodium leak in Superheater 2, in which a total of 40 DEGFs occurred in a short period of time, cast doubt on the choice of DBA for PFR and it was obvious that multiple tube failure incidents had to be considered. A revised Safety Case for PFR was constructed based on an event tree and is presented in this paper. Also, in this paper the engineering work carried out on the plant in order to reduce the frequency of occurrence of multiple tube failures and the R and D programme initiated to remove unnecessary pessimism from the postulated multiple tube failure incidents are described. (author). 2 refs, 16 figs, 1 tab

Determination of the concentration profile of chemical elements in superheater pipes

International Nuclear Information System (INIS)

Aldape U, F.; Aspiazu F, J.

1986-05-01

This work has for object to determine the profile of concentration of chemical elements at trace level in a superheater pipe of Thermoelectric Plants using the X-ray emission spectroscopy technique induced by protons coming from the Accelerator of the Nuclear Center. In the X-ray detection, a Si Li detector was used. The technique was chosen because it allows a multielemental analysis, of high sensitivity and precision. The results can help to understand the problems that are had in the change of flexibility or of corrosion. This will be from utility to the Federal Electricity Commission (CFE). (Author)

Prototype fast reactor steam generator unit pressure vessel repairs

International Nuclear Information System (INIS)

Daniels, B.D.; Green, D.; Henderson, J.D.C.

1993-01-01

The prototype fast reactor at Dounreay has experienced a number of unscheduled shutdowns due to leaking reheater and superheater shell welds. There was a need to determine the cracking mechanism and to design a general repair technique simultaneously. Detailed investigations revealed that the crack locations correlated with the positions of rectification welds made at the time of vessel manufacture. A creep crack growth mechanism was identified; this requires through wall residual stress for through cracks to develop. A repair technique has been devised and successfully applied to the sites of a number of leaks. (author)

Fireside corrosion of superheaters/reheaters in advanced power plants

Energy Technology Data Exchange (ETDEWEB)

Syed, A.U.; Simms, N.J.; Oakey, J.E. [Cranfield Univ. (United Kingdom). Energy Technology Centre

2010-07-01

The generation of increasing amounts of electricity while simultaneously reducing environmental emissions (CO{sub 2}, SO{sub 2}, NO{sub x} particles, etc) has become a goal for the power industry worldwide. Co-firing biomass and coal in new advanced pulverised fuel power plants is one route to address this issue, since biomass is regarded as a CO{sub 2} neutral fuel (i.e. CO{sub 2} uptake during its growth equals the CO{sub 2} emissions produced during its combustion) and such new advanced power plants operate at higher efficiencies than current plants as a result of using steam systems with high temperatures and pressures. However, co-firing has the potential to cause significant operational challenges for such power plants as amongst other issues, it will significantly change the chemistry of the deposits on the heat exchanger surfaces and the surrounding gas compositions. As a result these critical components can experience higher corrosion rates, and so shorter lives, causing increased operational costs, unless the most appropriate materials are selected for their construction. This paper reports the results of a series of 1000 hour laboratory corrosion tests that have been carried out in controlled atmosphere furnaces, to assess the effect of biomass/coal co-firing on the fireside corrosion of superheaters/reheaters. The materials used for the tests were one ferritic alloy (T92), two austenitic alloys (347HFG and HR3C) and one nickel based alloy (alloy 625). Temperatures of 600 and 650 C were used to represent the metal temperatures in advanced power plants. During these exposures, traditional mass change data were recorded as the samples were recoated with the simulated deposits. After these exposures, cross-sections through samples were prepared using standard metallographic techniques and then analysed using SEM/EDX. Pre-exposure micrometer and post-exposure image analyser measurements were used so that the metal wastage could be calculated. These data are

Failure Investigation of Radiant Platen Superheater Tube of Thermal Power Plant Boiler

Science.gov (United States)

Ghosh, D.; Ray, S.; Mandal, A.; Roy, H.

2015-04-01

This paper highlights a case study of typical premature failure of a radiant platen superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement and chemical analysis, are conducted as part of the investigations. Apart from these, metallographic analysis and fractography are also conducted to ascertain the probable cause of failure. Finally it has been concluded that the premature failure of the super heater tube can be attributed to localized creep at high temperature. The corrective actions has also been suggested to avoid this type of failure in near future.

Optimal Claiming Strategies in Bonus Malus Systems and Implied Markov Chains

Directory of Open Access Journals (Sweden)

Arthur Charpentier

2017-11-01

Full Text Available In this paper, we investigate the impact of the accident reporting strategy of drivers, within a Bonus-Malus system. We exhibit the induced modification of the corresponding class level transition matrix and derive the optimal reporting strategy for rational drivers. The hunger for bonuses induces optimal thresholds under which, drivers do not claim their losses. Mathematical properties of the induced level class process are studied. A convergent numerical algorithm is provided for computing such thresholds and realistic numerical applications are discussed.

Pengaruh Pajak, Tunneling Incentive dan Mekanisme Bonus Terhadap Keputusan Transfer Pricing

Directory of Open Access Journals (Sweden)

Mispiyanti Mispiyanti

2016-03-01

Full Text Available The aim of the research is to find empirical effidence of tax, tunneling incentive and bonus mecahnism toward transfer pricing decision taken by manufacturing companies listed in Indonesia Stock Exchange. The research population are manufacturing companies listed in Indonesia Stock Exchange arround 2010 to 2013. The samples were taken using purposive sampling method. The research results show that tax and bonus mechanism do not have effect toward companies’ transfer pricing decision. While, tunneling incentive has effect toward companies’transfer pricing decision.

BALCOFISH - a BONUS+ project in the Baltic Sea

DEFF Research Database (Denmark)

Strand, Jakob

The project BALCOFISH, acronym for "Integration of pollutant gene responses and fish ecology in Baltic coastal fisheries and management" is a newly started 3-years BONUS+-project funded by Baltic Organisations Network for Funding Science EEIG (www.bonusportal.org) with focus on contaminants...

Mitigating the Risk of Stress Corrosion of Austenitic Stainless Steels in Advanced Gas Cooled Reactor Boilers

International Nuclear Information System (INIS)

Bull, A.; Owen, J.; Quirk, G.; G, Lewis; Rudge, A.; Woolsey, I.S.

2012-09-01

Advanced Gas-Cooled Reactors (AGRs) operated in the UK by EDF Energy have once-through boilers, which deliver superheated steam at high temperature (∼500 deg. C) and pressure (∼150 bar) to the HP turbine. The boilers have either a serpentine or helical geometry for the tubing of the main heat transfer sections of the boiler and each individual tube is fabricated from mild steel, 9%Cr1%Mo and Type 316 austenitic stainless steel tubing. Type 316 austenitic stainless steel is used for the secondary (final) superheater and steam tailpipe sections of the boiler, which, during normal operation, should operate under dry, superheated steam conditions. This is achieved by maintaining a specified margin of superheat at the upper transition joint (UTJ) between the 9%Cr1%Mo primary superheater and the Type 316 secondary superheater sections of the boiler. Operating in this mode should eliminate the possibility of stress corrosion cracking of the Type 316 tube material on-load. In recent years, however, AGRs have suffered a variety of operational problems with their boilers that have made it difficult to maintain the specified superheat margin at the UTJ. In the case of helical boilers, the combined effects of carbon deposition on the gas side and oxide deposition on the waterside of the tubing have resulted in an increasing number of austenitic tubes operating with less than the specified superheat margin at the UTJ and hence the possibility of wetting the austenitic section of the boiler. Some units with serpentine boilers have suffered creep-fatigue damage of the high temperature sections of the boiler, which currently necessitates capping the steam outlet temperature to prevent further damage. The reduction in steam outlet temperature has meant that there is an increased risk of operation with less than the specified superheat margin at the UTJ and hence stress corrosion cracking of the austenitic sections of the boiler. In order to establish the risk of stress

The Australian baby bonus maternity payment and birth characteristics in Western Australia.

Directory of Open Access Journals (Sweden)

Kristjana Einarsdóttir

Full Text Available BACKGROUND: The Australian baby bonus maternity payment introduced in 2004 has been reported to have successfully increased fertility rates in Australia. We aimed to investigate the influence of the baby bonus on maternal demographics and birth characteristics in Western Australia (WA. METHODS AND FINDINGS: This study included 200,659 birth admissions from WA during 2001-2008, identified from administrative birth and hospital data-systems held by the WA Department of Health. We estimated average quarterly birth rates after the baby bonus introduction and compared them with expected rates had the policy not occurred. Rate and percentage differences (including 95% confidence intervals were estimated separately by maternal demographics and birth characteristics. WA birth rates increased by 12.8% following the baby bonus implementation with the greatest increase being in mothers aged 20-24 years (26.3%, 95%CI = 22.0,30.6, mothers having their third (1.6%, 95%CI = 0.9,2.4 or fourth child (2.2%, 95%CI = 2.1,2.4, mothers living in outer regional and remote areas (32.4%, 95%CI = 30.2,34.6, mothers giving birth as public patients (1.5%, 95%CI = 1.3,1.8, and mothers giving birth in public hospitals (3.5%, 95%CI = 2.6,4.5. Interestingly, births to private patients (-4.3%, 95%CI = -4.8,-3.7 and births in private hospitals (-6.3%, 95%CI = -6.8,-5.8 decreased following the policy implementation. CONCLUSIONS: The introduction of the baby bonus maternity payment may have served as an incentive for women in their early twenties and mothers having their third or fourth child and may have contributed to the ongoing pressure and staff shortages in Australian public hospitals, particularly those in outer regional and remote areas.

Review about corrosion of superheaters tubes in biomass plants

International Nuclear Information System (INIS)

Berlanga-Labari, C.; Fernandez-Carrasquilla, J.

2006-01-01

The design of new biomass-fired power plants with increased steam temperature raises concerns of high-temperature corrosion. The high potassium and chlorine contents in many biomass, specially in wheat straw, are potentially harmful elements with regard to corrosion. Chlorine may cause accelerated corrosion resulting in increased oxidation, metal wastage, internal attack, void formations and loose non-adherent scales. The most severe corrosion problems in biomass-fired systems are expected to occur due to Cl-rich deposits formed on superheater tubes. In the first part of this revision the corrosion mechanism proposed are described in function of the conditions and compounds involved. The second part is focused on the behaviour of the materials tested so far in the boiler and in the laboratory. First the traditional commercial alloys are studied and secondly the new alloys and the coasting. (Author). 102 refs

Production of A356 aluminum alloy wheels by thixo-forging combined with a low superheat casting process

Directory of Open Access Journals (Sweden)

Wang Shuncheng

2013-09-01

Full Text Available The A356 aluminum alloy wheels were produced by thixo-forging combined with a low superheat casting process. The as-cast microstructure, microstructure evolution during reheating and the mechanical properties of thixo-forged wheels made from the A356 aluminum alloy were studied. The results show that the A356 aluminum alloy round billet with fine, uniform and non-dendritic grains can be obtained when the melt is cast at 635 篊. When the round billet is reheated at 600 篊 for 60 min, the non-dendritic grains are changed into spherical ones and the round billet can be easily thixo-forged into wheels. The tensile strength, yield strength and elongation of the thixo-forged wheels with T6 heat treatment are 327.6 MPa, 228.3 MPa and 7.8%, respectively, which are higher than those of a cast wheel. It is suggested that the thixo-forging combined with the low superheat casting process is an effective technique to produce aluminum alloy wheels with high mechanical properties.

Incentivizing with Bonus in a College Statistics Course

Science.gov (United States)

Ingalls, Victoria

2018-01-01

Many studies have argued the negative effects of external rewards on internal motivation while others assert that external motivation does not necessarily undermine intrinsic motivation. At a private university, students were given the option to earn bonus points for achieving mastery in the online homework systems associated with Statistics and…

Sustainable bonuses: Sign of corporate responsibility or window dressing?

NARCIS (Netherlands)

Kolk, A.; Perego, P.

2014-01-01

Despite a strong plea for integrating sustainability goals into traditional corporate bonus schemes, a comprehensive implementation of these systems has been lacking until recently. This article explores four illustrative cases from the Netherlands, where several multinationals started to pioneer

Natural gas turbine topping for the iris reactor

International Nuclear Information System (INIS)

Oriani, L.; Lombardi, C.; Paramonov, D.

2001-01-01

Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

Natural gas turbine topping for the iris reactor

Energy Technology Data Exchange (ETDEWEB)

Oriani, L.; Lombardi, C. [Politecnico di Milano, Milan (Italy); Paramonov, D. [Westinghouse Electric Corp., LLC, Pittsburgh, PA (United States)

2001-07-01

Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

12 CFR 2.4 - Bonus and incentive plans.

Science.gov (United States)

2010-01-01

... plan based on the sale of credit life insurance if payments to the employee or officer in any one year... 12 Banks and Banking 1 2010-01-01 2010-01-01 false Bonus and incentive plans. 2.4 Section 2.4 Banks and Banking COMPTROLLER OF THE CURRENCY, DEPARTMENT OF THE TREASURY SALES OF CREDIT LIFE INSURANCE...

Description of reactor fuel breeding with three integral concepts

International Nuclear Information System (INIS)

Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

1979-01-01

The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

Stress corrosion cracking in superheater and reheater austenitic tubing

Energy Technology Data Exchange (ETDEWEB)

Dooley, R. Barry [Structural Integrity Associates, Inc., Charlotte, NC (United States); Bursik, Albert [PowerPlant Chemistry GmbH, Neulussheim (Germany)

2011-02-15

University 101 courses are typically designed to help incoming first-year undergraduate students to adjust to the university, develop a better understanding of the college environment, and acquire essential academic success skills. Why are we offering a special Boiler and HRSG Tube Failures PPChem 101? The answer is simple, yet very conclusive: - There is a lack of knowledge on the identification of tube failure mechanisms and for the implementation of adequate counteractions in many power plants, particularly at industrial power and steam generators. - There is a lack of knowledge to prevent repeat tube failures. The vast majority of BTF/HTF have been, and continue to be, repeat failures. It is hoped that the information about the failure mechanisms of BTF supplied in this course will help to put plant engineers and chemists on the right track. The major goal of this course is the avoidance of repeat BTF. This eights lesson is focused on Stress Corrosion Cracking in Superheater and Reheater Austenitic Tubing. (orig.)

Evaluation of materials' corrosion and chemistry issues for advanced gas cooled reactor steam generators using full scale plant simulations

International Nuclear Information System (INIS)

Woolsey, I.S.; Rudge, A.J.; Vincent, D.J.

1998-01-01

Advanced Gas Cooled Reactors (AGRS) employ once-through steam Generators of unique design to provide steam at approximately 530 degrees C and 155 bar to steam turbines of similar design to those of fossil plants. The steam generators are highly compact, and have either a serpentine or helical tube geometry. The tubes are heated on the outside by hot C0 2 gas, and steam is generated on the inside of the tubes. Each individual steam generator tube consists of a carbon steel feed and primary economiser section, a 9%Cr steel secondary economiser, evaporator and primary superheater, and a Type 316L austenitic stainless steel secondary superheater, all within a single tube pass. The multi-material nature of the individual tube passes, the need to maintain specific thermohydraulic conditions within the different material sections, and the difficulties of steam generator inspection and repair, have required extensive corrosion-chemistry test programmes to ensure waterside corrosion does not present a challenge to their integrity. A major part of these programmes has been the use of a full scale steam generator test facility capable of simulating all aspects of the waterside conditions which exist in the plant. This facility has been used to address a wide variety of possible plant drainage/degradation processes. These include; single- and two-phase flow accelerated corrosion of carbon steel, superheat margins requirements and the stress-corrosion behaviour of the austenitic superheaters, on-load corrosion of the evaporator materials, and iron transport and oxide deposition behaviour. The paper outlines a number of these, and indicates how they have been of value in helping to maintain reliable operation of the plant. (author)

Factors in the selection of broiler tube materials for a civil fast reactor

Energy Technology Data Exchange (ETDEWEB)

Tyzack, C; Chitty, A

1975-07-01

This paper briefly considers some of the factors which must be balanced in the selection of a boiler tube material for a Civil Fast Reactor. The merits and possible demerits of low alloy ferritic steels and the austenitic Alloy 800 are compared with respect to waterside corrosion resistance, mechanical properties, fabrication and weldability and possible effects of exposure to the sodium environment under normal and fault conditions. It is pointed out that although there is operational experience of most of the materials in boiler superheater applications there is little or none in evaporative regimes. (author)

Microscopical investigation of steamside oxide on X20CrMoV121 superheater tubes

DEFF Research Database (Denmark)

Montgomery, Melanie; Hansson, Anette Nørgaard; Jensen, Søren A.

2011-01-01

X20CrMoV121 is a 12%Cr martensitic steel which has been used in power plants in Europe for many decades. Superheater tubes exposed for various durations up to 135,000 hours in power plants in Denmark at steam temperatures varying from 450 to 575°C were investigated. Light optical and scanning ele...... electron microscopy was used to investigate steamside oxide morphologies. At all temperatures there is a double layered oxide, however, the inner:outer oxide thickness is not always equal. At the lower steam temperature range of...

29 CFR 778.503 - Pseudo “percentage bonuses.”

Science.gov (United States)

2010-07-01

... such a scheme is artificially low, and the difference between the wages paid at the hourly rate and the... Regulations Relating to Labor (Continued) WAGE AND HOUR DIVISION, DEPARTMENT OF LABOR STATEMENTS OF GENERAL... part, a true bonus based on a percentage of total wages—both straight time and overtime wages—satisfies...

29 CFR 778.208 - Inclusion and exclusion of bonuses in computing the “regular rate.”

Science.gov (United States)

2010-07-01

... 29 Labor 3 2010-07-01 2010-07-01 false Inclusion and exclusion of bonuses in computing the... Inclusion and exclusion of bonuses in computing the “regular rate.” Section 7(e) of the Act requires the inclusion in the regular rate of all remuneration for employment except seven specified types of payments...

Effect of reinforcement amount, mold temperature, superheat, and mold thickness on fluidity of in-situ Al-Mg2Si composites

Directory of Open Access Journals (Sweden)

Reza Vatankhah Barenji

2018-01-01

Full Text Available In the present study, the effects of mold temperature, superheat, mold thickness, and Mg2Si amount on the fluidity of the Al-Mg2Si as-cast in-situ composites were investigated using the mathematical models. Composites with different amounts of Mg2Si were fabricated, and the fluidity and microstructure of each were then analyzed. For this purpose, the experiments were designed using a central composite rotatable design, and the relationship between parameters and fluidity were developed using the response surface method. In addition, optical and scanning electron microscopes were used for microstructural observation. The ANOVA shows that the mathematical models can predict the fluidity accurately. The results show that by increasing the mold temperature from 25 °C to 200 °C, superheat from 50 °C to 250 °C, and thickness from 3 mm to 12 mm, the fluidity of the composites decreases, where the mold thickness is more effective than other factors. In addition, the higher amounts of Mg2Si in the range from 15wt.% to 25wt.% lead to the lower fluidity of the composites. For example, when the mold temperature, superheat, and thickness are respectively 100 °C, 150 °C, and 7 mm, the fluidity length is changed in the range of 11.9 cm to 15.3 cm. By increasing the amount of Mg2Si, the morphology of the primary Mg2Si becomes irregular and the size of primary Mg2Si is increased. Moreover, the change of solidification mode from skin to pasty mode is the most noticeable microstructural effect on the fluidity.

Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

Directory of Open Access Journals (Sweden)

Mohammed Al Hajri

2015-04-01

Full Text Available Dissimilar metal weld (DMW joint between alloyed steel (AS and stainless steel (SS failed at one of intermediate temperature superheater (ITSH tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results of the investigation point out the limitation of Carbides precipitation at the alloyed steel/welding interface. This is synonym of creep stage I involvement in the failure of ITSH. Improper post-welding operation and bending moment are considered as root causes of the premature failure.

Deterministic economic analysis of feedlot Red Angus young steers: slaughter weights and bonus

Directory of Open Access Journals (Sweden)

Paulo Santana Pacheco

2015-03-01

Full Text Available The joint analysis of indicators of the investment project is very relevant in making decisions, resulting in more consistent information regarding risk assessment and its confrontation with the possibility of return. This research aimed to evaluate the economic feasibility of Red Angus young steers finished in feedlot, slaughtered at 340, 373, 396 or 430kg with use of various financial indicators, marketed with or without bonus. The purchase of feeder cattle and feeding were variable costs with a higher share in the total cost. In the analysis with bonus, the regression analysis to gross margin, net margin, net present value, benefit:cost index and additional return on investment showed quadratic behavior, with the point of maximum at 406kg (R$ 185.17, 406kg (R$ 161.76 , 393kg (R$ 128.29, 392kg (1.12, 392kg (11.98%, respectively. In the analysis without bonus, gross margin and net margin showed a quadratic response (346kg, with R$ 110.31 and R$ 86.90, respectively, while for the other indicators, there was a linear reduction as an increase in slaughter weight.

Re-evaluation of superheat conditions postulated in NRC Information Notice 84-90

International Nuclear Information System (INIS)

Alsammarae, A.; Kruger, D.; Beutel, D.; Spisak, M.

1994-01-01

Information Notice 84-90, ''Main Steam Line Break Effect on Environmental Qualification of Equipment,'' describes a potential problem regarding existing plant analyses and Equipment Qualification (EQ) related to a postulated Main Steam Line Break (MSLB) with releases of superheated stream. This notice states that certain methodologies for computing mass and energy releases for a postulated MSLB did not account for heat transfer from the steam generator tube bundles if they were uncovered. Due to this potential change in the original environmental analysis, the EQ of various components may not consider the thermal environment which could result from superheated steam. Subsequent technical assessments may determine that the existing qualification basis for equipment/components does not envelop the postulated superheat condition. Corrective actions need to be taken to demonstrate that the affected equipment is qualified

[Participant structure and economic benefit of prevention bonus programmes in company health insurance funds].

Science.gov (United States)

Friedrichs, M; Friedel, H; Bödeker, W

2009-10-01

This study investigates differences in sex, age, and educational level between participants and non-participants of prevention bonus programmes. The differences in the utilisation of drugs, hospital care, and sickness absence before the start of the programmes between these groups are also shown. Finally the economic benefit of the health insurance funds attributed to these programmes is estimated. Data from some 5.2 million insured subjects of 74 company health insurance funds in Germany were linked to information on enrollment into a prevention bonus programme anonymously. In a descriptive analysis the differences in the sociodemographic patterns between both groups are shown as well as the differences in costs to the health insurances in the three sectors mentioned above. The benefit to the health insurance funds is estimated by means of an analysis of covariance. Prevention bonus programmes yields an annual benefit of at least 129 euro per participant. Men aged 40 and older and women aged 30 and older are more likely to opt into such a programme. The same is true for persons with a higher educational level. There are only few differences in health-care utilisation between the participants and non-participants of the programmes before enrollment. Only 1.4% of all insured persons participated in the programmes. There is at least a short-term gain to both involved parties: the insured and the health insurance funds. The programmes are not dominated by deadweight effects. Long-term effects and effectiveness of prevention bonus programmes still have to be investigated. Copyright Georg Thieme Verlag KG Stuttgart . New York.

The intrinsic inferiority of efficiency wages to damages and conditional bonuses

NARCIS (Netherlands)

Geest, Gerrit de; Dari Mattiacci, G.; Siegers, J.J.

2004-01-01

In this paper, we argue that, as an enforcement mechanism, efficiency wages are intrinsically inferior to damages and to conditional bonuses an alternative positive sanction system overlooked in the labor economics literature, under which rents are only paid if monitoring has effectively taken

Investigation into Cause of High Temperature Failure of Boiler Superheater Tube

Science.gov (United States)

Ghosh, D.; Ray, S.; Roy, H.; Shukla, A. K.

2015-04-01

The failure of the boiler tubes occur due to various reasons like creep, fatigue, corrosion and erosion. This paper highlights a case study of typical premature failure of a final superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement, chemical analysis, oxide scale thickness measurement, microstructural examination are conducted as part of the investigations. Apart from these investigations, sulfur print, Energy Dispersive spectroscopy (EDS) and X ray diffraction analysis (XRD) are also conducted to ascertain the probable cause of failure of final super heater tube. Finally it has been concluded that the premature failure of the super heater tube can be attributed to the combination of localized high tube metal temperature and loss of metal from the outer surface due to high temperature corrosion. The corrective actions have also been suggested to avoid this type of failure in near future.

The Intrinsic Inferiority of Efficiency Wages to Damages and Conditional Bonuses

NARCIS (Netherlands)

de Geest, G.G.A.; Dari Mattiacci, G.; Siegers, J.J.

In this paper, we argue that, as an enforcement mechanism, efficiency wages are intrinsically inferior to damages and to conditional bonuses – an alternative positive sanction system overlooked in the labor economics literature, under which rents are only paid if monitoring has effectively taken

The Impact of Higher Fixed Pay and Lower Bonuses on Productivity

NARCIS (Netherlands)

Bun, M.J.G.; Huberts, L.C.E.

This study analyzes the effects of performance related pay on productivity exploiting a change in the payment structure of a large Dutch marketing company. Specifically, we investigate the consequences for company sales of higher fixed pay in combination with lower bonuses. Exploiting shift level

Development of heat transfer models for gap cooling

Energy Technology Data Exchange (ETDEWEB)

Kohriyama, Tamio; Murase, Michio; Tamaki, Tomohiko [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

In a severe accident of a light water reactor (LWR), heat transfer models in a narrow annular gap between superheated core debris and a reactor pressure vessel (RPV) are important to evaluate the integrity of RPV and emergency procedures. This paper discusses the effects of superheat on the heat flux based on existing data. In low superheat conditions, the heat flux in the narrow gap is higher than the heat flux in pool nucleate boiling due to restricted flow area. It approaches the nucleate boiling heat flux as superheat increasing and reaches a critical value subject to the counter-current flow limiting (CCFL) at the top end of the gap. A heat transfer correlation was derived as a function of dimensionless superheat and a Kutateladze-type CCFL correlation was deduced for critical heat flux (CHF) restricted by CCFL, which gave good prediction for a wide range of the CHF data. Effect of an angle of inclination of the gap could also be incorporated in the CCFL correlation. In high superheat conditions, the heat flux in the narrow gap maintains a similar shape to the pool boiling curve but shifts the position to a higher superheated side than the pool boiling except film boiling, which could be expressed by the typical pool film boiling correlation. Incorporating quench test data, the heat flux correlation was derived as a function of dimensionless superheat using the same formula for the low superheat and the Kutateladze-type CCFL correlation was deduced for CHF. The CHF at the high superheat was 3-4 times as large as CHF at the low superheat and this difference was well predicted by different flow patterns in the gap and the balance of pressure gradients between gas and liquid phases. (author)

Effects of Degree of Superheat on the Running Performance of an Organic Rankine Cycle (ORC Waste Heat Recovery System for Diesel Engines under Various Operating Conditions

Directory of Open Access Journals (Sweden)

Kai Yang

2014-04-01

Full Text Available This study analyzed the variation law of engine exhaust energy under various operating conditions to improve the thermal efficiency and fuel economy of diesel engines. An organic Rankine cycle (ORC waste heat recovery system with internal heat exchanger (IHE was designed to recover waste heat from the diesel engine exhaust. The zeotropic mixture R416A was used as the working fluid for the ORC. Three evaluation indexes were presented as follows: waste heat recovery efficiency (WHRE, engine thermal efficiency increasing ratio (ETEIR, and output energy density of working fluid (OEDWF. In terms of various operating conditions of the diesel engine, this study investigated the variation tendencies of the running performances of the ORC waste heat recovery system and the effects of the degree of superheat on the running performance of the ORC waste heat recovery system through theoretical calculations. The research findings showed that the net power output, WHRE, and ETEIR of the ORC waste heat recovery system reach their maxima when the degree of superheat is 40 K, engine speed is 2200 r/min, and engine torque is 1200 N·m. OEDWF gradually increases with the increase in the degree of superheat, which indicates that the required mass flow rate of R416A decreases for a certain net power output, thereby significantly decreasing the risk of environmental pollution.

Modeling and Performance of Bonus-Malus Systems: Stationarity versus Age-Correction

Directory of Open Access Journals (Sweden)

Søren Asmussen

2014-03-01

Full Text Available In a bonus-malus system in car insurance, the bonus class of a customer is updated from one year to the next as a function of the current class and the number of claims in the year (assumed Poisson. Thus the sequence of classes of a customer in consecutive years forms a Markov chain, and most of the literature measures performance of the system in terms of the stationary characteristics of this Markov chain. However, the rate of convergence to stationarity may be slow in comparison to the typical sojourn time of a customer in the portfolio. We suggest an age-correction to the stationary distribution and present an extensive numerical study of its effects. An important feature of the modeling is a Bayesian view, where the Poisson rate according to which claims are generated for a customer is the outcome of a random variable specific to the customer.

DINAMIKA PENGANGGURAN TERDIDIK: TANTANGAN MENUJU BONUS DEMOGRAFI DI INDONESIA

Directory of Open Access Journals (Sweden)

Sri Maryati

2015-07-01

Full Text Available According to the United Nations demographic transition that occurred in recent decades in Indonesia would be an opportunity for Indonesia to reach a demographic dividend in the period 2020-2030 . By the time the productive age population amounted to twice that of the non - productive population . These opportunities should be best utilized as it will only happen once and it can happen if the population of working age have a job and sufficient income . thus this demographic bonus can actually stimulate the economy of Indonesia in the future . But on the other hand, Indonesia is currently facing serious problems of labor that is still large numbers of educated unemployment . The number of unemployed educated annually feared will continue to grow as the number of college graduates also continue to grow , but not all college graduates can be accommodated in the workplace , consequently leads to an increase in the number of educated unemployed . The main purpose of this study is to analyze the dynamics of educated unemployment in Indonesia and the steps that need to be done by the government and people of Indonesia in order to face the era of demographic bonus, so it does not become a wave of mass unemployment, particularly educated unemployment in Indonesia.

Building Integrated PV and PV/Hybrid Products - The PV:BONUS Experience: Preprint

Energy Technology Data Exchange (ETDEWEB)

Thomas, H.; Pierce, L. K.

2001-10-01

Presented at the 2001 NCPV Program Review Meeting: Successes and lessons learned from PV:BONUS (Building Opportunities in the United States in PV). This program has funded the development of PV or PV/hybrid products for building applications.

Risk:reward sharing contracts in the oil industry: the effects of bonus:penalty schemes

International Nuclear Information System (INIS)

Kemp, A.G.; Stephen, L.

1999-01-01

Partnering and alliancing among oil companies and their contractors have become common in the oil industry in recent years. The risk:reward mechanisms established very often incorporate bonus/penalty schemes in relation to agreed base values. This paper examines the efficiency requirements of such schemes. The effects of project cost and completion risks on the risk: reward positions of field investors and contractors with and without bonus/penalty schemes are examined with the aid of Monte Carlo simulation analysis. The schemes increase the total risk for contractors and have consequence for their cost of capital and optimal risk-bearing arrangements within the industry. (author)

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

DECISION SUPPORT SYSTEM PEMBERIAN BONUS TAHUNAN PADA KARYAWAN BERDASARKAN KINERJA KARYAWAN MENGGUNAKAN METODE SIMPLE ADDITIVE WEIGHTING (STUDY KASUS : STMIK PRINGSEWU

Directory of Open Access Journals (Sweden)

Zulkifli Zulkifli

2017-05-01

Full Text Available Bonuses is one method that is widely used as a form of tribute to workers whose performance so far can be considered satisfactory by the company. So is the case with STMIK Pringsewu that rewards Her staff as a token of appreciation for its performance over the years. However, the annual bonus is only given to employees who are considered berprestasit. It required a decision support system (DSS or decesion support system that can take into account all the criteria that support and to help facilitate the decision making process. This decision support system using Simple Additive Wighting (SAW. The issue of decision support is basically a form of election of the various alternative actions that may be selected include discipline or the number of absences in a year, length of employment, crafts, and work in a year that the process through specific mechanisms, in hopes of generating a best decision. Employees who got the votes of 100% working receive an annual bonus that is the departure of the hajj, work evaluation 87.5% earn an annual bonus that is the departure of Umrah, and the assessment work 75% earn an annual bonus that is the departure of the general allowance.

Damage distribution and remnant life assessment of a super-heater outlet header used for long time

Energy Technology Data Exchange (ETDEWEB)

Hiroyuki, Okamura [Science Univ. of Tokyo (Japan); Ryuichi, Ohotani [Kyoto Univ. (Japan); Kazuya, Fujii [Japan Power Engineering and Inspection Corp., Tokyo (Japan); Masashi, Nakashiro; Fumio, Takemasa; Hideo, Umaki; Tomiyasu, Masumura [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

1998-11-01

This paper presents the results of investigation on evaluating damage distribution to base metals and welded joints in the thickness direction and evaluate damage on ligaments. Thick wall tested sample was the superheater outlet header component long term serviced in high pressure and temperature condition in thermal power plant. The simulate unused steel of component material was made from sample by suitable heat treatment, and the extent of damage was assessed based on a comparison of nondestructive and destructive test results between simulate unused and aged samples. Damage evaluation was also made by FEM structural stress analysis. (orig./MM)

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

Lecture background notes on transient sodium boiling and voiding in fast reactors

International Nuclear Information System (INIS)

Okrent, D.; Fauske, H.K.

1972-01-01

This set of lecture background notes includes the following: (1) Introductory remarks on fast reactor safety, which are intended to provide some perspective on the role played by sodium boiling. (2) A discussion of superheat which reviews the experimental data and nucleation models with emphasis on the pressure-temperature history effect on radius of active cavity sites, including the role played by inert gas. (3) A discussion of the growth and collapse of spherical bubbles. (4) A historical description of the development of computer codes to describe voiding and a detailed description of the analytical formulation of typical models for calculating voiding due to boiling, fission gas release, and molten fuel-coolant interaction. (U.S.)

The EBR-II steam generating system - operation, maintenance, and inspection

International Nuclear Information System (INIS)

Buschman, H.W.; Penney, W.H.; Longua, K.J.

2002-01-01

The Experimental Breeder Reactor II (EBR-II) has operated for 20 years at the Idaho National Engineering Laboratory near Idaho Falls. EBR-II is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. EBR-II has operated at a capacity factor over 70% in the past few years. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C and 8.62 MPa. The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. During the 20 years of operation, components of the steam generator have been subjected to a variety of inspections including visual, dimensional, and ultrasonic. One superheater was removed from service because of anomalous performance and was replaced with an evaporator which was removed, examined, and converted into a superheater. Overall operating experience of the system has been excellent and essentially trouble free. Inspections have not revealed any conditions that are performance or life limiting. (author)

MOTIVASI BONUS, PAJAK, DAN UTANG DALAM TINDAKAN MANAJEMEN LABA (STUDI PERUSAHAAN MANUFAKTUR YANG TERDAFTAR DI BURSA EFEK INDONESIA PERIODE 2013-2015

Directory of Open Access Journals (Sweden)

Novia Fitri Kusumawardani

2017-10-01

Full Text Available The objective of the empirical study is to examine and analyze the influence of Bonus Motivation, Tax Motivation, and Debt Motivation in Earning Management. The variables studied are bonus, tax, debt, firm size, and earning management. The sample of this study is the manufacture company that listed in Indonesia Stock Exchange in 2013 until 2015. The sample used in this study were 165 samples. The classical assumption used is normality test, multicollinearity, autocorrelation, and heteroscedasticity test. Hypothesis test used is the coefficient of determination (R ², simultant test (F-test and invidual test(t-test. The results of this study indicate that (1 Bonus Motivation has positive significant influence in earning management. (2 Tax Motivation has positive significant influence in earning management (3 Debt Motivation has positive significant influence in earning management. Keywords: Earning Management, Bonus, Income Tax, Debt to Assets Ratio, Firm Size.

The creep life of superheater and reheater tubes under varying pressure conditions in operational boilers

International Nuclear Information System (INIS)

Mizen, D.C.; Plastow, B.

1975-01-01

The first of each manufacturer's 500 MW boilers supplied to the CEGB (Central Electricity Generating Board) have been subjected to an extensive programme of tests for performance optimization and safe operation. Around 250 thermocouples on superheater and reheater tubes have in each case been monitored as part of the exercise. The readings are corrected and used to compute creep rupture damage based on internationally agreed stress rupture data and a simple cumulative damage concept. Comparison of the design creep rupture life and the cumulative life consumed has in several applications been invaluable in influencing operating procedures and arranging tube modifications or replacements, so that loss of generation by creep rupture failure is minimized. (author)

Controlling Banker's Bonuses: Efficient Regulation or Politics of Envy?

OpenAIRE

Matthews, Kent; Matthews, Owen

2009-01-01

The positive relationship between bank CEO compensation and risk taking is a well established empirical fact. The global banking crisis has resulted in a chorus of demands to control banker’s bonuses and thereby curtail their risk taking activities in the hope that the world can avoid a repeat in the future. However, the positive relationship is not a causative one. In this paper we argue that the cushioning of banks downside risks provide the incentive for banks to take excessive risk and de...

An empirical examination of negotiated goals and performance-to-goal following the introduction of an incentive bonus plan with participative goal setting

NARCIS (Netherlands)

Anderson, S.W.; Dekker, H.C.; Sedatole, K.L.

2010-01-01

Prior research documents performance improvements following the implementation of pay-for-performance (PFP) bonus plans. However, bonus plans typically pay for performance relative to a goal, and the manager whose performance is to be evaluated often participates in setting the goal. In these

Bonus systems and their effects on safety: an interview-based pilot study at the Swedish nuclear power plants

International Nuclear Information System (INIS)

Torbioern, Ingemar; Mattson, Malin

2009-03-01

The aim of this pilot study has been to describe and analyse potential effects on safety-related behaviour and risks associated with the bonus systems currently used at Swedish nuclear plants. To this end and in order to establish a frame of reference several theories on motivation were consulted regarding the relevance of monetary rewards. In addition empirical evidence on effects upon behaviours in general and safety behaviours in particular was taken into consideration, as well as a systems and a rationalist perspective on organisations. The resulting frame of reference was used for a descriptive mapping of the bonus systems and for the formulation of a semi-structured interview schedule intended to capture the experiences of those concerned by the systems. A total of 15 interviews were performed with staff of different functions and organisational positions. Results of the study do not indicate any negative effects on safety-related behaviours. Rather they indicate that safety-behaviours may be promoted insofar as bonus rewards are linked to performance goals concerning safety. All of the bonus-systems may be characterised as low in incentive intensity, i.e. produce small effects on motivation and performance. Still, as the systems differ in design and in the way they are perceived, they also represent different challenges in order to function more efficiently as parameters

Degradation of superheater tubes made of austenitic T321H steel after long term service

Energy Technology Data Exchange (ETDEWEB)

Hernas, Adam [Silesian Technical Univ., Katowice (Poland). Faculty of Material Science; Augustyniak, Boleslaw; Chmielewski, Marek [Gdansk Univ. of Technology (Poland). Mechanical Dept.; Sablik, M.J. [Applied Magnetic and Physical Modeling, LLC, San Antonio, TX (United States)

2010-07-01

There are presented results of complementary tests performed for the evaluation of creep damage in austenitic steel grade T321H exploited over 200,000 hours in the secondary superheater part of a power plant boiler. The following techniques have been applied: SEM microscopy, X-ray diffraction, tensile tests, hardness measurements and novel eddy current inspection. The novel eddy current inspection is proposed as a non-destructive method of estimating the creep damage stage of austenite steel boiler tubes after long-term service in power plants. We compare the results provided by the different techniques and discuss the correlations and also point out the problems which need to be addressed in order to elaborate the remaining life assessment of austenitic boiler tubes. (orig.)

CADDS [Computer-aided Drafting and Design System] brings quality and precision to the Canadian Maple [research reactor

International Nuclear Information System (INIS)

Goland, D.

1989-01-01

Atomic Energy of Canada Ltd (AECL) has found that using the ''intelligent'' Computer-Aided Drafting and Design System (CADDS) helped address design problems at an early stage and led to productivity gains of around 50 per cent. Other bonuses were the quality and precision of the designs and documents produced. Its application to the MAPLE research reactor project is described. (author)

Effects of Different Fuel Specifications and Operation Conditions on the Performance of Coated and Uncoated Superheater Tubes in Two Different Biomass-Fired Boilers

DEFF Research Database (Denmark)

Wu, Duoli; Dahl, Kristian V.; Madsen, Jesper L.

2018-01-01

Fireside corrosionis a serious concern in biomass firing powerplants such that the efficiency of boilers is limited by high temperature corrosion. Application of protective coatings on superheater tubes is a possible solution to combat fireside corrosion. The current study investigates the corros......Fireside corrosionis a serious concern in biomass firing powerplants such that the efficiency of boilers is limited by high temperature corrosion. Application of protective coatings on superheater tubes is a possible solution to combat fireside corrosion. The current study investigates...... the corrosion performance of coated tubes compared to uncoated Esshete 1250 and TP347H tubes, which were exposed in two different biomass-fired boilers for one year. Data on the fuel used, temperature of the boilers, and temperature fluctuations are compared for the two boilers, and how these factors influence...... deposit formation, corrosion, and the stability of the coatings is discussed. The coatings (Ni and Ni2Al3) showed protective behavior ina wood-fired plant where the outlet steam temperature was 520 °C. However, at the plant that fired straw with an outlet steam temperature of 540 °C and where severe...

Adding a Performance-Based Component to Surface Warfare Officer Bonuses: Will it Affect Retention?

National Research Council Canada - National Science Library

Carman, Aron S; Mudd, Ryan M

2008-01-01

... Authorization and the current officer inventory beginning at 9 years of commissioned service. The objective of this study was to analyze the 13-year retention effect of adding a performance-based component to the SWO Critical Skills Bonus (CSB...

CFD analysis of temperature imbalance in superheater/reheater region of tangentially coal-fired boiler

Science.gov (United States)

Zainudin, A. F.; Hasini, H.; Fadhil, S. S. A.

2017-10-01

This paper presents a CFD analysis of the flow, velocity and temperature distribution in a 700 MW tangentially coal-fired boiler operating in Malaysia. The main objective of the analysis is to gain insights on the occurrences in the boiler so as to understand the inherent steam temperature imbalance problem. The results show that the root cause of the problem comes from the residual swirl in the horizontal pass. The deflection of the residual swirl due to the sudden reduction and expansion of the flow cross-sectional area causes velocity deviation between the left and right side of the boiler. This consequently results in flue gas temperature imbalance which has often caused tube leaks in the superheater/reheater region. Therefore, eliminating the residual swirl or restraining it from being diverted might help to alleviate the problem.

The impact of instant reward programs and bonus premiums on consumer purchase behavior

NARCIS (Netherlands)

Minnema, Alec; Bijmolt, Tammo H. A.; Non, Marielle C.

This study examines the impact of an instant reward program (IRP) with bonus premiums on consumer purchase behavior. An IRP is a rapidly growing form of short-term program that rewards consumers instantly with small premiums per fixed spending, where these premiums are part of a larger set of

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1997-12-01

In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

Test of New Readout Electronics for the BONuS12 Experiment

Energy Technology Data Exchange (ETDEWEB)

Ehrhart, Mathieu [Inst. de Physique Nucleaire (IPN), Orsay (France)

2017-07-01

For decades, electron-proton scattering experiments have been providing a large amount of data on the proton structure function. However, because of the instability of free neutrons, fewer experiments have been able to study the neutron structure function. The BONuS collaboration at Jefferson Laboratory addresses this challenge by scattering electrons off a deuterium target, using a RTPC capable of detecting the low-momentum spectator protons near the target. Events of electrons scattering on almost free neutrons are selected by constraining the spectator protons to very low momenta and very backward scattering angles. In 2005, BONuS successfully measured the neutron structure with scattering electrons of up to 5.3 GeV energy. An extension of this measurement has been approved using the newly upgraded 12 GeV electron beam and CLAS12 (CEBAF Large Acceptance Spectrometer). For this new set of measurements, a new RTPC detector using GEM trackers is being developed to allow measurements of spectator protons with momenta as low as 70 MeV/c. The new RTPC will use a new readout electronic system, which is also used by other trackers in CLAS12. This thesis will present the first tests of this electronics using a previously built RTPC of similar design.

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

Energy Technology Data Exchange (ETDEWEB)

Kjellstroem, B; Larsson, A E

1967-04-15

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

International Nuclear Information System (INIS)

Kjellstroem, B.; Larsson, A.E.

1967-04-01

In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

Controlling banker’s bonuses: Efficient regulation or politics of envy?

OpenAIRE

Matthews, Kent; Mattehws, Owen; Cardiff University

2009-01-01

The positive relationship between bank CEO compensation and risk taking is a well established empirical fact. The global banking crisis has resulted in a chorus of demands to control banker’s bonuses and thereby curtail their risk taking activities in the hope that the world can avoid a repeat in the future. However, the positive relationship is not a causative one. In this paper we argue that the cushioning of banks downside risks provide the incentive for banks to take excessive risk and de...

Bonus Point System for Refuse Classification and Sustainable Development: A Study in China

Directory of Open Access Journals (Sweden)

Shijie Guo

2017-09-01

Full Text Available The rapid growth of household waste not only endangers the environment and people’s health, but also limits social and economic development. The effective sorting and recycling of garbage can control this problem. Adopting a semi-quantitative case study method, our researchers investigated the effect of a bonus point system for refuse classification that improves the accuracy of refuse classification and the residents’ environment awareness. In the system, residents will receive some gifts after sorting the garbage correctly. We also investigated the attitudes of residents and companies towards this novel system. Our researchers employed various methods to analyze garbage-sorting data, questionnaires completed by residents, and interview records. The results show that use of a bonus point system affects the management of domestic waste by improving the accuracy and enhancing the awareness of garbage sorting. Overall, residents support the system and benefit from it, which increases participation and consciousness of environmental protection. However, continuous publicity and coordination of various policies are required to promote the wide-range implementation and sustainable development of this system.

The steam generating heavy water reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

THE STUDY OF INDONESIA’S READINESS TO COPE WITH DEMOGRAPHIC BONUS: A REVIEW OF POPULATION LAW

Directory of Open Access Journals (Sweden)

Edie Toet Hendratno

2015-09-01

Full Text Available The shift from a population mapping to investigate the population bonus on demographic aspect enhances the economic value this study might contribute for. Indonesian population as the fourth largest number requires some policies to cope with the millennium challenges. Improving nationwide database maintains government on-going strategy to manage its population that is expected to serve all stakeholders for any quest towards economic development. The study uses mixed method with the explanatory sequential strategy. The qualitative approach is used, using social network analysis, supported by desk study, in-depth interview, focus group discussion, and literature studies. This study attempts to provide for improvement to the extant regulations on identity card, mostly electronic-KTP. Demographic bonus is an interesting topic given that the younger the population, the higher the employment demand rate will be. Having systematic database bank and access is expected to alleviate the challenge on high population growth rate in Indonesia. Besides, recommendations are addressed to the policy maker (the Government, mostly on the review or amendment of the extant regulations that might not in line with the database systematic improvements. The study is a primary thesis from a review of population law, using multidisciplinary approach, i.e. population economics (demography, legal study, and public policy that can be used as a testing basis to answer further demographic bonus from an exploration of other scientific inquiries.

Study on superheat of TiAl melt during cold crucible levitation melting. TiAl no cold crucible levitation yokai ni okeru yoto kanetsudo no kenkyu

Energy Technology Data Exchange (ETDEWEB)

Miwa, K.; Kobayashi, K.; Ninomiya, M. (Government Industrial Research Institute, Nagoya, Nagoya (Japan))

1992-06-20

Investigations were given on effects of test sample weights and sample positions in cold crucibles on superheat of melts when the intermetallic compound TiAl is melted using cold crucible levitation melting process, one of noncontaminated melting processes. The cold crucibles used in the experiment are a water-cooled copper crucible with an inner diameter of 42 mm and a length of 140 mm, into which a column-like ingot sample with an outer diameter of 32 mm (Al containing Ti at 33.5% by mass) was put and melted using the levitation melting. Comparisons and discussions were given on the relationship between sample weights and melt temperatures, the relationship between positions of the inserted samples and melt temperatures, and the state of contamination at melting of casts obtained from the melts resulted from the levitation melting and high-frequency melting poured into respective ceramic dies. Elevating the superheat temperature of the melts requires optimizing the sample weights and positions. Melt temperatures were measured using a radiation thermometer and a thermocouple, and the respective measured values were compared. 7 refs., 4 figs., 1 tab.

Failure Analysis and Magnetic Evaluation of Tertiary Superheater Tube Used in Gas-Fired Boiler

Science.gov (United States)

Mohapatra, J. N.; Patil, Sujay; Sah, Rameshwar; Krishna, P. C.; Eswarappa, B.

2018-02-01

Failure analysis was carried out on a prematurely failed tertiary superheater tube used in gas-fired boiler. The analysis includes a comparative study of visual examination, chemical composition, hardness and microstructure at failed region, adjacent and far to failure as well as on fresh tube. The chemistry was found matching to the standard specification, whereas the hardness was low in failed tube compared to the fish mouth opening region and the fresh tube. Microscopic examination of failed sample revealed the presence of spheroidal carbides of Cr and Mo predominantly along the grain boundaries. The primary cause of failure is found to be localized heating. Magnetic hysteresis loop (MHL) measurements were carried out to correlate the magnetic parameters with microstructure and mechanical properties to establish a possible non-destructive evaluation (NDE) for health monitoring of the tubes. The coercivity of the MHL showed a very good correlation with microstructure and mechanical properties deterioration enabling a possible NDE technique for the health monitoring of the tubes.

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1996-08-01

In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

Family policy in the Czech Republic: Redistribution of wealth through the child tax bonus

Directory of Open Access Journals (Sweden)

Robert Jahoda

2013-01-01

Full Text Available Families with children are traditionally the target group of the social system in developed countries. This paper deals with one component of family policy in the Czech Republic, which is household entitlement. The main focus is on the child tax bonus (hereafter CTB. The paper is divided into descriptive and methodological-analytical parts. The descriptive section provides basic information about the beneficiaries of CTB. In the latter section we formulate research questions about the impacts and effects of CTB. We discover that the influence of tax instruments has grown in recent years. The amount of the tax bonus for children exceeded CZK 3 billion in 2009, with almost 22% of all households with children eligible. Although CTB is income-tested, its redistributive impact is rather small – approximately 80% of recipients cannot be considered as poor. Outcomes from our microsimulation model reveal that 82 to 86% households with CTB were at the same time modelled as eligible and therefore we can use microsimulation techniques for future analyses of policy change.

Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

International Nuclear Information System (INIS)

Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

1965-11-01

The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

Dual turbine power plant and method of operating such plant, especially one having an HTGR steam supply

International Nuclear Information System (INIS)

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

1977-01-01

A power plant including dual steam turbine-generators connected to pass superheat and reheat steam from a steam generator which derives heat from the coolant gas of a high temperature gas-cooled nuclear reactor is described. Associated with each turbine is a bypass line to conduct superheat steam in parallel with a high pressure turbine portion, and a bypass line to conduct superheat steam in parallel with a lower pressure turbine portion. Auxiliary steam turbines pass a portion of the steam flow to the reheater of the steam generator and drive gas blowers which circulate the coolant gas through the reactor and the steam source. Apparatus and method are disclosed for loading or unloading a turbine-generator while the other produces a steady power output. During such loading or unloading, the steam flows through the turbine portions are coordinated with the steam flows through the bypass lines for protection of the steam generator, and the pressure of reheated steam is regulated for improved performance of the gas blowers. 33 claims, 5 figures

Secondary flows in the cooling channels of the high-performance light-water reactor

Energy Technology Data Exchange (ETDEWEB)

Laurien, E.; Wintterle, Th. [Stuttgart Univ., Institute for Nuclear Technolgy and Energy Systems (IKE) (Germany)

2007-07-01

The new design of a High-Performance Light-Water Reactor (HPLWR) involves a three-pass core with an evaporator region, where the compressed water is heated above the pseudo-critical temperature, and two superheater regions. Due to the strong dependency of the supercritical water density on the temperature significant mass transfer between neighboring cooling channels is expected if the temperature is unevenly distributed across the fuel element. An inter-channel flow is then superimposed to the secondary flow vortices induced by the non-isotropy of turbulence. In order to gain insight into the resulting flow patterns as well as into temperature and density distributions within the various subchannels of the fuel element CFD (Computational Fluid Dynamics) calculations for the 1/8 fuel element are performed. For simplicity adiabatic boundary conditions at the moderator box and the fuel element box are assumed. Our investigation confirms earlier results obtained by subchannel analysis that the axial mass flux is significantly reduced in the corner subchannel of this fuel element resulting in a net mass flux towards the neighboring subchannels. Our results provide a first estimation of the magnitude of the secondary flows in the pseudo-critical region of a supercritical light-water reactor. Furthermore, it is demonstrated that CFD is an efficient tool for investigations of flow patterns within nuclear reactor fuel elements. (authors)

PEMBERIAN BONUS KEPADA PEKERJA DI PT. SANG HYANG SERI (PERSERO DALAM RANGKA MENINGKATKAN PRODUKTIVITAS KERJA DITINJAU DARI UNDANG-UNDANG NOMOR 13 TAHUN 2003 TENTANG KETENAGAKERJAAN

Directory of Open Access Journals (Sweden)

Rani Apriani, S.E., S.H., M.H.

2016-05-01

Full Text Available Workers primary motivation to do a job is to earn wages as agreed on the contract, but sometimes, the wages has some issue, one of which is when thethe worker felt that they already produce the job exceed the standard worker. Bonus classified as non-wage income provided by the company in order to encourage workers to be more discipline, industrious, productive, and increase the workers earning in PT. Sang Hyang Seri. The purpose of this research was to examine the rule used by PT. Sang Hyang Seri when giving bonuses to increase productivity and to examine the bonus implementation in PT. Sang Hyang Seri comply with Act Number 3 on 2013 related with employment.

Incentivizing Advanced Mathematics Study at Upper Secondary Level: The Case of Bonus Points in Ireland

Science.gov (United States)

Treacy, Páraic Thomas

2018-01-01

Secondary level mathematics education in Ireland has recently experienced a period of significant change with the introduction of new curricula and the addition of an incentive to study upper secondary mathematics at the most advanced level (Higher Level). This incentive, typically referred to as 'bonus points', appears to have aided a significant…

Failure problems in superheater spacers of steam generators; Problematica de fallas en espaciadores de sobrecalentadores de generadores de vapor

Energy Technology Data Exchange (ETDEWEB)

Chacon Nava, Jose G; Martinez Villafane, Alberto [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Fuentes Samaniego, Raul [Universidad Autonoma de Nuevo Leon (Mexico); Mojica Calderon, Cecilio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1988-12-31

In this article the general aspects of the steam generator superheater fixed spacers failures are analyzed, emphasis is made on the influence several aspects such as the operation of the unit have, the appropriate execution of welds and the selection of binding materials. Likewise several recommendations are made to bring the failures to a minimum. [Espanol] En este articulo se analizan aspectos generales de fallas en espaciadores fijos de sobrecalentadores de generadores de vapor, y se hace hincapie en la influencia que tienen diversos aspectos tales como la operacion de la unidad, la adecuada ejecucion de soldaduras y la seleccion del material de aporte. Asimismo, se proponen algunas recomendaciones para reducir al minimo las fallas.

Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

International Nuclear Information System (INIS)

Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

1981-01-01

Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

The Cost Effectiveness of the U.S. Export Enhancement Program Bonus Allocation Mechanism

OpenAIRE

Humei Wang; Richard J. Sexton

2004-01-01

The U.S. Export Enhancement Program is evaluated from the perspective of the cost effectiveness of its bonus allocation mechanism. The current mechanism resembles a discriminatory-price, common-value auction. However, auction theory suggests that a discriminatory auction may not be optimal in this setting for several reasons. This article evaluates the current format relative to an alternative, uniform-price auction. Estimation results reveal evidence of strategic bidder behavior under the cu...

Pre-oxidation and its effect on reducing high-temperature corrosion of superheater tubes during biomass firing

DEFF Research Database (Denmark)

Okoro, Sunday Chukwudi; Kvisgaard, M.; Montgomery, Melanie

2017-01-01

Superheater tubes in biomass-fired power plants experience high corrosion rates due to condensation of corrosive alkali chloride-rich deposits. To explore the possibility of reducing the corrosion attack by the formation of an initial protective oxide layer, the corrosion resistance of pre......-oxidised Al and Ti-containing alloys (Kanthal APM and Nimonic 80A, respectively) was investigated under laboratory conditions mimicking biomass firing. The alloys were pre-oxidised at 900°C for 1 week. Afterwards, pre-oxidised samples, and virgin non-pre-oxidised samples as reference, were coated...... with a synthetic deposit of KCl and exposed at 560°C for 1 week to a gas mixture typical of biomass firing. Results show that pre-oxidation could hinder the corrosion attack; however, the relative success was different for the two alloys. While corrosion attack was observed on the pre-oxidised Kanthal APM, the pre...

Operating experience of the EBR-II steam generating system

International Nuclear Information System (INIS)

Buschman, H.W.; Penney, W.H.; Quilici, M.D.; Radtke, W.H.

1981-01-01

The Experimental Breeder Reactor II (EBR-II) is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C (820 F) and 8.62 MPa (1250 psi). The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. Safety and reliability are maximized by using duplex tubes and tubesheets. The performance of the system has been excellent and essentially trouble free. The operating experience of EBR-II provides confidence that the technology can be applied to commercial LMFBR's for an abundant supply of energy for the future. 5 refs

20 CFR 670.620 - Are Job Corps students eligible to receive cash allowances and performance bonuses?

Science.gov (United States)

2010-04-01

... Student Support § 670.620 Are Job Corps students eligible to receive cash allowances and performance... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false Are Job Corps students eligible to receive cash allowances and performance bonuses? 670.620 Section 670.620 Employees' Benefits EMPLOYMENT AND...

Critical heat fluxes in tubular fuel elements of nuclear power reactors

International Nuclear Information System (INIS)

Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

1974-01-01

The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

Floating market bonus in the Renewable Energy Law. Opportunity or risk for renewable energies?; Gleitende Marktpraemie im EEG. Chance oder Risiko fuer die Erneuerbaren?

Energy Technology Data Exchange (ETDEWEB)

Nestle, Uwe [Buendnis 90/die Gruenen, Flensburg (Germany). Bundesarbeitsgemeinschaft Energie

2011-03-15

The call for market integration amid the discussion about the future development of the Renewable Energy Law (EEG) can no longer be overheard. One instrument frequently mentioned in this context is the so-called ''floating market bonus'', which the German Federal Government is considering to introduce, as it declares in its Energy Concept. On weighing the pros and cons of a floating market bonus it becomes clear that the goals it is intended to achieve such as adjusting feed-in to demand cannot be brought about in this way. At the same time it poses a menace to the continued rapid deployment of renewable energy production plants for electricity generation. It would therefore be better turn to other instruments that are better suited for achieving the purpose of the floating market bonus.

Incentives in statutory health insurance bonus schemes - Communication as an underrated precondition of success

OpenAIRE

Viviane Scherenberg; Gerd Glaeske

2009-01-01

Aim - Bonus schemes within German statutory health insurance (GKV) use monetary incentives to promote health-conscious behaviour, particularly amongst risk groups. The idea is to exploit a latent potential for participation in money-saving preventive measures. First studies suggest that incidental effects (good risks) are more common than prevention effects. The purpose of the article is to present factors contributing to the successfulness of incentive schemes. Methods - To outline the findi...

Bonus algorithm for large scale stochastic nonlinear programming problems

CERN Document Server

Diwekar, Urmila

2015-01-01

This book presents the details of the BONUS algorithm and its real world applications in areas like sensor placement in large scale drinking water networks, sensor placement in advanced power systems, water management in power systems, and capacity expansion of energy systems. A generalized method for stochastic nonlinear programming based on a sampling based approach for uncertainty analysis and statistical reweighting to obtain probability information is demonstrated in this book. Stochastic optimization problems are difficult to solve since they involve dealing with optimization and uncertainty loops. There are two fundamental approaches used to solve such problems. The first being the decomposition techniques and the second method identifies problem specific structures and transforms the problem into a deterministic nonlinear programming problem. These techniques have significant limitations on either the objective function type or the underlying distributions for the uncertain variables. Moreover, these ...

WWC Review of the Report "A Big Apple for Educators: New York City's Experiment with Schoolwide Performance Bonuses. Final Evaluation Report." What Works Clearinghouse Single Study Review

Science.gov (United States)

What Works Clearinghouse, 2013

2013-01-01

The study examined in this paper focuses on whether monetary bonuses for teachers improved schoolwide academic achievement in New York City public schools. Study authors analyzed data from 389 high-need elementary, middle, and high schools in New York City in the first year of the bonus program (2007-08) and from 371 of those same schools in the…

An Analysis of the Effect of the U. S. Marine Corps' Lump Sum Selective Reenlistment Bonus Program on Reenlistment Decisions

National Research Council Canada - National Science Library

Barry, Robert

2001-01-01

... the impact of personal characteristics, civilian pay, unemployment, and the lump sum bonus on reenlistment decisions, Marine retention probabilities under the lump sum payment program are compared...

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1998-06-01

In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

A computational approach for thermomechanical fatigue life prediction of dissimilarly welded superheater tubes

Energy Technology Data Exchange (ETDEWEB)

Krishnasamy, Ram-Kumar; Seifert, Thomas; Siegele, Dieter [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany)

2010-07-01

In this paper a computational approach for fatigue life prediction of dissimilarly welded superheater tubes is presented and applied to a dissimilar weld between tubes made of the nickel base alloy Alloy617 tube and the 12% chromium steel VM12. The approach comprises the calculation of the residual stresses in the welded tubes with a multi-pass dissimilar welding simulation, the relaxation of the residual stresses in a post weld heat treatment (PWHT) simulation and the fatigue life prediction using the remaining residual stresses as initial condition. A cyclic fiscoplasticity model is used to calculate the transient stresses and strains under thermocyclic service loadings. The fatigue life is predicted with a damage parameter which is based on fracture mechanics. The adjustable parameters of the model are determined based on LCF and TMF experiments. The simulations show, that the residual stresses that remain after PWHT further relax in the first loading cycles. The predicted fatigue lives depend on the residual stresses and, thus, on the choice of the loading cycle in which the damage parameter is evaluated. It the first loading cycle, where residual stresses are still present, is considered, lower fatigue lives are predicted compared to predictions considering loading cycles with relaxed residual stresses. (orig.)

Corrosion protection on superheaters of waste to energy plants. Experience with material and application; Korrosionsschutz im Ueberhitzerbereich. Erfahrungen mit Werkstoff und Applikation aus Qualitaetsbegleitungen

Energy Technology Data Exchange (ETDEWEB)

Schmidl, Werner; Herzog, Thomas; Magel, Gabi; Mueller, Wolfgang; Spiegel, Wolfgang [CheMin GmbH, Augsburg (Germany)

2011-07-01

Corrosion induced by chlorine at high temperatures and corrosion by salt melts sometimes cause severe risk and loss of operational availability in waste- and biomass-fired power plants. This corrosion very often affects the superheater. Due to high maintenance needs, several approaches to anti-corrosion coating have been developed. Nickel-based alloys such as alloy 625 are chosen to be applied as cladding or by thermal spraying. Operation periods have been considerably increased by these methods. But still there are some shortcomings in corrosion protection due to application and/or material. (orig.)

Determination of the concentration profile of chemical elements in superheater pipes; Determinacion del perfil de concentracion de elementos quimicos en tubos de sobrecalentadores

Energy Technology Data Exchange (ETDEWEB)

Aldape U, F; Aspiazu F, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1986-05-15

This work has for object to determine the profile of concentration of chemical elements at trace level in a superheater pipe of Thermoelectric Plants using the X-ray emission spectroscopy technique induced by protons coming from the Accelerator of the Nuclear Center. In the X-ray detection, a Si Li detector was used. The technique was chosen because it allows a multielemental analysis, of high sensitivity and precision. The results can help to understand the problems that are had in the change of flexibility or of corrosion. This will be from utility to the Federal Electricity Commission (CFE). (Author)

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

International Nuclear Information System (INIS)

Carey, W.M.; Gavin, A.P.; Bobis, J.P.; Sheen, S.H.; Anderson, T.T.; Doolittle, R.D.; Albrecht, R.W.

1977-01-01

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapour bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapour bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low ( 0 C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary. (author)

Phase identification and internal stress analysis of steamside oxides on superheater tubes by means of X-ray diffraction

Energy Technology Data Exchange (ETDEWEB)

Pantleon, Karen; Montgomery, Melanie [Technical Univ. of Denmark, Lyngby (Denmark). Inst. of Manufacturing Engineering and Management

2005-05-01

For superheater tubes, the adherence of the inner steamside oxide is especially important as spallation of this oxide results in a) blockage of loops which cause insufficient steam flow through the superheaters and subsequently overheating and tube failure and b) spalled oxide can cause erosion of turbine blades. Oxide spallation is a serious problem for austenitic steels where the significant differences of the thermal expansion coefficients of steel and oxide cause relatively high thermal stresses. Usually, various oxides layered within the scale are suggested from microscopical observations of the morphology and/or topography of the oxide scale accompanied by the analysis of chemical elements present. Reports about the application of X-ray diffraction on the study of steamside oxide formation are very scarce in literature. If applied at all, XRD-studies are restricted to ideally flat samples oxidized under laboratory conditions, but relation to real operating conditions and the effect of the real sample geometry is missing. Within the frame of the project, steamside oxides on plant exposed components of ferritic/ martensitic X20CrMoV12-1 as well as fine- and coarse-grained austenitic TP347H were studied by means of X-ray diffraction. Depth dependent phase analysis on sample segments cut from the tubes was carried out by means of grazing incidence diffraction and, in order to obtain information from a larger depth, conventional XRD was combination with stepwise mechanical removal of the steamside oxides. After each removal step phase analysis was performed both on the segments and on the removed powders. Phase specific stress analysis was carried out on rings cut from the tube. Results show that steamside oxides on X20CrMoV12-1 consist of pure Hematite at the surface followed by a relatively thick layer of pure Magnetite. Both phases are under relatively high tensile stresses. While the phase composition of the Hematite layer appears to be the same for all

Three applications of a bonus relation for gravity amplitudes

International Nuclear Information System (INIS)

Spradlin, Marcus; Volovich, Anastasia; Wen, Congkao

2009-01-01

Arkani-Hamed et al. have recently shown that all tree-level scattering amplitudes in maximal supergravity exhibit exceptionally soft behavior when two supermomenta are taken to infinity in a particular complex direction, and that this behavior implies new non-trivial relations amongst amplitudes in addition to the well-known on-shell recursion relations. We consider the application of these new 'bonus relations' to MHV amplitudes, showing that they can be used quite generally to relate (n-2)!-term formulas typically obtained from recursion relations to (n-3)!-term formulas related to the original BGK conjecture. Specifically we provide (1) a direct proof of a formula presented by Elvang and Freedman, (2) a new formula based on one due to Bedford et al., and (3) an alternate proof of a formula recently obtained by Mason and Skinner. Our results also provide the first direct proof that the conjectured BGK formula, only very recently proven via completely different methods, satisfies the on-shell recursion.

Performance Evaluation of a Printed Circuit Steam Generator for Integral Reactors: A Feasibility Test

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han-Ok; Yoon, Juhyeon; Kim, Young In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of); Seo, Jang-won; Choi, Brain [Alfa Laval Korea Ltd., Daejeon (Korea, Republic of)

2015-05-15

SMART (System-integrated Modular Advanced ReacTor) is a small-sized integral type pressurized water reactor. It adopts advanced design features such as structural safety improvement, system simplification, and component modularization to achieve highly enhanced safety and improved economics. The design issues related to further safety enhancement and cost reduction have received significant attention to increase its competitiveness in the global small reactor market. For the cost reduction, it is important to design the reactor vessel as small as possible. Thus, it is necessary to reduce the volume of main components such as a steam generator. Its manufacturing processes of the chemical etching and diffusion bonding provide high effectiveness, high compactness, and inherent structural safety under high temperatures and high pressures. Thus, it is expected to be an alternative to the conventional shell and tube type steam generator in SMART. In this paper, simple thermal-hydraulic performance measurement of a small-scale printed circuit steam generator (PCSG) is conducted to investigate the feasibility of applying it to SMART. The simple thermal-hydraulic performance of the PCSG has been experimentally evaluated. A small-scale PCHE is employed to investigate the feasibility of operating it as a steam generator. The performance assessment reveals that the PCSG stably produces superheated steam, and the increased degree of superheat is obtained at lower water flow rate. However, the flow instability is increased with the decrease of the water flow rate. Thus, it is required to apply the orifice design into the cold side plate to suppress the density-wave oscillations. The pressure drops and heat transfer rates increase with the water flow rate.

A Big Apple for Educators: New York City's Experiment with Schoolwide Performance Bonuses. Final Evaluation Report. Monograph

Science.gov (United States)

Marsh, Julie A.; Springer, Matthew G.; McCaffrey, Daniel F.; Yuan, Kun; Epstein, Scott; Koppich, Julia; Kalra, Nidhi; DiMartino, Catherine; Peng, Art

2011-01-01

In the 2007-2008 school year, the New York City Department of Education and the United Federation of Teachers jointly implemented the Schoolwide Performance Bonus Program in a random sample of the city's high-needs public schools. The program lasted for three school years, and its broad objective was to improve student performance through…

Fluidized bed combustion of refuse-derived fuel in presence of protective coal ash

Energy Technology Data Exchange (ETDEWEB)

Ferrer, Eduardo [CIRCE, Universidad de Zaragoza, Maria de Luna, 3, Zaragoza (Spain); Aho, Martti [VTT Processes, P.O. Box 1603, 40101 Jyvaeskylae (Finland); Silvennoinen, Jaani; Nurminen, Riku-Ville [Kvaerner Power, P.O.Box 109, FIN-33101 Tampere (Finland)

2005-12-15

Combustion of refuse-derived fuel (RDF) alone or together with other biomass leads to superheater fouling and corrosion in efficient power plants (with high steam values) due to vaporization and condensation of alkali chlorides. In this study, means were found to raise the portion of RDF to 40% enb without risk to boilers. This was done by co-firing RDF with coal and optimizing coal quality. Free aluminum silicate in coal captured alkalies from vaporized alkali chlorides preventing Cl condensation to superheaters. Strong fouling and corrosion were simultaneously averted. Results from 100 kW and 4 MW CFB reactors are reported. (author)

NCS-1 dependent learning bonus and behavior outputs of self-directed exploration

Science.gov (United States)

Mun, Ho-Suk

Animals explore a new environment and learn about their surroundings. "Exploration" refers to all activities that increase the information obtained from an animal. For this study, I determined a molecule that mediates self-directed exploration, with a particular focus on rearing behavior and vocalization. Rearing can be either self-directed exploration or escape-oriented exploration. Self-directed exploration can be driven by the desire to gather information about environments while escape-oriented exploration can be driven by fear or anxiety. To differentiate between these two concepts, I compared rearing and other behaviors in three different conditions 1) novel dim (safe environment), which induces exploration based rearing; 2) novel bright (fearful environment), which elicits fear driven rearing; and 3) familiar environment as a control. First, I characterized the effects on two distinct types of environment in exploratory behavior and its effect on learning. From this, I determined that self-directed exploration enhances spatial learning while escape-oriented exploration does not produce a learning bonus. Second, I found that NCS-1 is involved in exploration, as well as learning and memory, by testing mice with reduced levels of Ncs-1 by point mutation and also siRNA injection. Finally, I illustrated other behavior outputs and neural substrate activities, which co-occurred during either self-directed or escape-oriented exploration. I found that high-frequency ultrasonic vocalizations occurred during self-directed exploration while low-frequency calls were emitted during escape-oriented exploration. Also, with immediate early gene imaging techniques, I found hippocampus and nucleus accumbens activation in self-directed exploration. This study is the first comprehensive molecular analysis of learning bonus in self-directed exploration. These results may be beneficial for studying underlying mechanisms of neuropsychiatric disease, and also reveal therapeutic

Four stream breakup of molten IFR [Integral Fast Reactor] metal fuel in sodium

International Nuclear Information System (INIS)

Gabor, J.D.; Purviance, R.T.; Aeschlimann, R.W.; Spencer, B.W.

1988-01-01

Tests have been conducted in which the breakup behavior of kilogram quantities of molten uranium, uranium-zirconium alloy, and uranium-iron alloy pour streams in 600C sodium was studied. A sodium depth of less than 0.3 m was required for hydrodynamic breakup and freezing of 25-mm pour streams of uranium and uranium-zirconium alloy with up to 400C melt superheat. The breakup material was primarily in the form of filaments and sheets with a settled bed voidage on the order of 0.9. The uranium-iron alloy with 800C melt superheat exhibited similar behavior except a sodium depth somewhat greater than 0.3 m was required for breakup and freezing of the particles

Operation of the water-to-sodium leak detection system at the experimental breeder reactor II

International Nuclear Information System (INIS)

Osterhout, M.M.

1978-01-01

A water-to-sodium leak detection system was installed at the Experimental Breeder Reactor II in April 1975. The system is designed for early detection of steam generator leaks, using hydrogen meters at the sodium outlets of the evaporators and superheaters. The leak detectors operate by measuring the rate of diffusion of hydrogen from the liquid sodium through a nickel membrane into a dynamic vacuum system. The advantages of this detection system are rapid response time, high sensitivity, stability, and reliability. The system was operated on an experimental basis for the first two years. During this period, data were obtained on detector stability, reliability, maintenance needs, computer interface requirements, calibration, and background hydrogen-level fluctuations. A generic defect in the original detectors was also discovered, requiring redesign of the units. When the new units were installed and proven to be reliable, the system was made fully operational. The data from the hydrogen meters are now used as the primary basis for detection of water-to-sodium leaks

Review about corrosion of superheaters tubes in biomass plants; Revision sobre la corrosion de tubos sobrecalentadores en plantas de biomasa

Energy Technology Data Exchange (ETDEWEB)

Berlanga-Labari, C.; Fernandez-Carrasquilla, J.

2006-07-01

The design of new biomass-fired power plants with increased steam temperature raises concerns of high-temperature corrosion. The high potassium and chlorine contents in many biomass, specially in wheat straw, are potentially harmful elements with regard to corrosion. Chlorine may cause accelerated corrosion resulting in increased oxidation, metal wastage, internal attack, void formations and loose non-adherent scales. The most severe corrosion problems in biomass-fired systems are expected to occur due to Cl-rich deposits formed on superheater tubes. In the first part of this revision the corrosion mechanism proposed are described in function of the conditions and compounds involved. The second part is focused on the behaviour of the materials tested so far in the boiler and in the laboratory. First the traditional commercial alloys are studied and secondly the new alloys and the coasting. (Author). 102 refs.

Device for starting a steam generator by heating sodium in a reactor

International Nuclear Information System (INIS)

Nakano, Hisao.

1975-01-01

Object: To enhance cooperation between ventilation and steam conditions of turbine and ventilation condition relative to a superheater at the time of starting a plant using a fast breeder, and to enhance safety with respect to failure of heat transmission tubes at the time of start. Structure: In a device in which steam generated in an evaporator is fed to a high pressure turbine through a super-heater and an outlet steam of high pressure turbine is reheated by means of a re-heater and fed into a turbine on the side of low pressure to drive the turbine for power generation, opening and closing valves are mounted on outlet and inlet pipes, respectively, of the heat transmission pipe in the super heater, said outlet and inlet pipes being connected by a bypass pipe. Upstream side of the opening and closing valve on the inlet pipe and the downstream side of the opening and closing valve on the outlet pipe and connected by a bypass pipe in the re-heater and said bypass pipe in the re-heater is provided with a steam heat exchanger to be heated by steam in the outlet of the superheater, and a steam line in an auxiliary boiler is connected to the side of re-heater from the opening and closing valve on the heat transmission pipe in the re-heater. (Hanada, M.)

Superheater fireside corrosion mechanisms in MSWI plants: Lab-scale study and on-site results

Energy Technology Data Exchange (ETDEWEB)

Brossard, J.M.; Chaucherie, X.; Nicol, F. [Veolia Environnement R and D, Zone Portuaire de Limay, 291 Avenue Dreyfous Ducas, Limay 78520 (France); Diop, I. [Veolia Environnement R and D, Zone Portuaire de Limay, 291 Avenue Dreyfous Ducas, Limay 78520 (France); Institut Jean Lamour, departement Chimie et physique des solides et des surfaces, UMR 7198 CNRS - Universite Henri Poincare Nancy 1, Vandoeuvre-Les-Nancy (France); Rapin, C.; Vilasi, M. [Institut Jean Lamour, departement Chimie et physique des solides et des surfaces, UMR 7198 CNRS - Universite Henri Poincare Nancy 1, Vandoeuvre-Les-Nancy (France)

2011-06-15

Combustion of municipal waste generates highly corrosive gases (HCl, SO{sub 2}, NaCl, KCl, and heavy metals chlorides) and ashes containing alkaline chlorides and sulfates. Currently, corrosion phenomena are particularly observed on superheater's tubes. Corrosion rates depend mainly on installation design, operating conditions i.e., gas and steam temperature and velocity of the flue gas containing ashes. This paper presents the results obtained using an innovative laboratory-scale corrosion unit, which simulates MSWI (Municipal Solid Waste Incineration) boilers conditions characterized by a temperature gradient at the metal tube in the presence of corrosive gases and ashes. The presented corrosion tests were realized on carbon steel at fixed metal temperature (400 C). The influence of the flue gas temperature, synthetic ashes composition, and flue gas flow pattern were investigated. After corrosion test, cross sections of tube samples were characterized to evaluate thickness loss and estimate corrosion rate while the elements present in corrosion layers were analyzed. Corrosion tests were carried out twice in order to validate the accuracy and reproducibility of results. First results highlight the key role of molten phase related to the ash composition and flue gas temperature as well as the deposit morphology, related to the flue gas flow pattern, on the mechanisms and corrosion rates. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

International Nuclear Information System (INIS)

Khartabil, H.F.

2000-01-01

A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

Structuring a Multiproduct Sales Quota-Bonus Plan for a Heterogeneous Sales Force: A Practical Model-Based Approach

OpenAIRE

Murali K. Mantrala; Prabhakant Sinha; Andris A. Zoltners

1994-01-01

This paper presents an agency theoretic model-based approach that assists sales managers in determining the profit-maximizing structure of a common multiproduct sales quota-bonus plan for a geographically specialized heterogeneous sales force operating in a repetitive buying environment. This approach involves estimating each salesperson's utility function for income and effort and using these models to predict individual sales achievements and the associated aggregate profit for the firm und...

Rapid evaporation at the superheat limit of methanol, ethanol, butanol and n-heptane on platinum films supported by low-stress SiN membranes.

Science.gov (United States)

Ching, Eric J; Avedisian, C Thomas; Cavicchi, Richard C; Chung, Do Hyun; Rah, Jeff; Carrier, Michael J

2016-10-01

The bubble nucleation temperatures of several organic liquids (methanol, ethanol, butanol, n-heptane) on stress-minimized platinum (Pt) films supported by SiN membranes is examined by pulse-heating the membranes for times ranging from 1 µs to 10 µs. The results show that the nucleation temperatures increase as the heating rates of the Pt films increase. Measured nucleation temperatures approach predicted superheat limits for the smallest pulse times which correspond to heating rates over 10 8 K/s, while nucleation temperatures are significantly lower for the longest pulse times. The microheater membranes were found to be robust for millions of pulse cycles, which suggests their potential in applications for moving fluids on the microscale and for more fundamental studies of phase transitions of metastable liquids.

Teacher Reactions to the Performance-Based Bonus Program: How the Expectancy Theory Works in the South Korean School Culture

Science.gov (United States)

Ha, Bong-Woon; Sung, Youl-Kwan

2011-01-01

This study was conducted in order to examine how and to what extent the implementation of the performance-based bonus program in South Korean schools has motivated teachers to improve their behavior, as well as to identify any other positive or negative effects of the program. Interviews with teachers indicated that a large percentage of teachers…

Prediction of mass fraction of agglomerated debris in a LWR severe accident

International Nuclear Information System (INIS)

Kudinov, P.; Davydov, M.

2011-01-01

Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation and solidification in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can create agglomerated debris and “cake” regions that increase hydraulic resistance of the bed and affect coolability of the bed. This paper discusses development and application of a conservative-mechanistic approach to quantify mass fractions of agglomerated debris. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with high superheat of binary oxidic simulant material melt is used for validation of the methods. Application of the approach to plant accident analysis suggests that melt superheat has less significant influence on agglomeration of the debris than jet penetration depth. The paper also discusses the impact of the uncertainty in the jet disintegration and penetration behavior on the agglomeration mode map. (author)

RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

International Nuclear Information System (INIS)

Blanchat, T.; Hassan, Y.

1989-01-01

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Energy Technology Data Exchange (ETDEWEB)

Huang, Jun-Lin; Zhou, Ke-Yi, E-mail: boiler@seu.edu.cn; Xu, Jian-Qun [Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education, School of Energy and Environment, Southeast University, Nanjing 210096, Jiangsu Province (China); Wang, Xin-Meng; Tu, Yi-You [School of Materials Science and Engineering, Southeast University, Nanjing 210096, Jiangsu Province (China)

2014-07-28

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Science.gov (United States)

Huang, Jun-Lin; Zhou, Ke-Yi; Wang, Xin-Meng; Tu, Yi-You; Xu, Jian-Qun

2014-07-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

International Nuclear Information System (INIS)

Huang, Jun-Lin; Zhou, Ke-Yi; Xu, Jian-Qun; Wang, Xin-Meng; Tu, Yi-You

2014-01-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

26 CFR 1.381(c)(11)-1 - Contributions to pension plan, employees' annuity plans, and stock bonus and profit-sharing plans.

Science.gov (United States)

2010-04-01

... 26 Internal Revenue 4 2010-04-01 2010-04-01 false Contributions to pension plan, employees... TAXES Insolvency Reorganizations § 1.381(c)(11)-1 Contributions to pension plan, employees' annuity... or transferor corporation in respect of any pension, annuity, stock bonus, or profit-sharing plan. (b...

Comparison between two rheocasting processes of damper cooling tube method and low superheat casting

Directory of Open Access Journals (Sweden)

Zhang Xiaoli

2014-09-01

Full Text Available To produce a high quality semisolid slurry that consists of fine primary particles uniformly suspended in the liquid matrix for rheoforming, chemical refining and electromagnetic or mechanical stirring are the two methods commonly used. But these two methods either contaminate the melt or incur high cost. In this study, the damper cooling tube (DCT method was designed to prepare semisolid slurry of A356 aluminum alloy, and was compared with the low superheat casting (LSC method - a conventional process used to produce casting slab with equiaxed dendrite microstructure for thixoforming route. A series of comparative experiments were performed at the pouring temperatures of 650 °C, 638 °C and 622 °C. Metallographic observations of the casting samples were carried out using an optical electron microscope with image analysis software. Results show that the microstructure of semisolid slurry produced by the DCT process consists of spherical primary α-Al grains, while equiaxed grains microstructure is found in the LSC process. The lower the pouring temperature, the smaller the grain size and the rounder the grain morphology in both methods. The copious nucleation, which could be generated in the DCT, owing to the cooling and stirring effect, is the key to producing high quality semisolid slurry. DCT method could produce rounder and smaller α-Al grains, which are suitable for semisolid processing; and the equivalent grain size is no more than 60 μm when the pouring temperature is 622 °C.

Rooftop PV system. PV:BONUS Phase 3B, final technical report

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-11-01

Under the PV:BONUS Program, ECD and United Solar developed, demonstrated and commercialized two new lightweight, flexible BIPV modules specifically designed as replacements for conventional asphalt shingles and standing seam metal roofing. These modules can be economically and aesthetically integrated into new residential and commercial buildings, and can be used to address the even larger roofing-replacement market. An important design feature of these modules, which minimizes the installation and balance-of-system costs, is their ability to be installed by conventional roofing contractors without special training. The modules are fabricated from high-efficiency, triple-junction spectrum-splitting a-Si alloy solar cells developed by ECD and United Solar. These cells are produced on thin, flexible stainless steel substrates and encapsulated with polymer materials. The Phase 3 program began in August 1995. The principal tasks and goals of this program, which have all been successfully completed by ECD and United Solar, are described in the body and appendices of this report.

Bonus systems and their effects on safety: an interview-based pilot study at the Swedish nuclear power plants; Bonussystem och dess inverkan paa saekerheten: en intervjubaserad pilotstudie vid de svenska kaernkraftverken

Energy Technology Data Exchange (ETDEWEB)

Torbioern, Ingemar; Mattson, Malin [Inst. of Psychology, Stockholm Univ., Stockholm (Sweden)

2009-03-15

The aim of this pilot study has been to describe and analyse potential effects on safety-related behaviour and risks associated with the bonus systems currently used at Swedish nuclear plants. To this end and in order to establish a frame of reference several theories on motivation were consulted regarding the relevance of monetary rewards. In addition empirical evidence on effects upon behaviours in general and safety behaviours in particular was taken into consideration, as well as a systems and a rationalist perspective on organisations. The resulting frame of reference was used for a descriptive mapping of the bonus systems and for the formulation of a semi-structured interview schedule intended to capture the experiences of those concerned by the systems. A total of 15 interviews were performed with staff of different functions and organisational positions. Results of the study do not indicate any negative effects on safety-related behaviours. Rather they indicate that safety-behaviours may be promoted insofar as bonus rewards are linked to performance goals concerning safety. All of the bonus-systems may be characterised as low in incentive intensity, i.e. produce small effects on motivation and performance. Still, as the systems differ in design and in the way they are perceived, they also represent different challenges in order to function more efficiently as parameters

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

Energy Technology Data Exchange (ETDEWEB)

Yoder, G.L.

2005-10-03

This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

New gene evolution in the bonus-TIF1-γ/TRIM33 family impacted the architecture of the vertebrate dorsal-ventral patterning network.

Science.gov (United States)

Wisotzkey, Robert G; Quijano, Janine C; Stinchfield, Michael J; Newfeld, Stuart J

2014-09-01

Uncovering how a new gene acquires its function and understanding how the function of a new gene influences existing genetic networks are important topics in evolutionary biology. Here, we demonstrate nonconservation for the embryonic functions of Drosophila Bonus and its newest vertebrate relative TIF1-γ/TRIM33. We showed previously that TIF1-γ/TRIM33 functions as an ubiquitin ligase for the Smad4 signal transducer and antagonizes the Bone Morphogenetic Protein (BMP) signaling network underlying vertebrate dorsal-ventral axis formation. Here, we show that Bonus functions as an agonist of the Decapentaplegic (Dpp) signaling network underlying dorsal-ventral axis formation in flies. The absence of conservation for the roles of Bonus and TIF1-γ/TRIM33 reveals a shift in the dorsal-ventral patterning networks of flies and mice, systems that were previously considered wholly conserved. The shift occurred when the new gene TIF1-γ/TRIM33 replaced the function of the ubiquitin ligase Nedd4L in the lineage leading to vertebrates. Evidence of this replacement is our demonstration that Nedd4 performs the function of TIF1-γ/TRIM33 in flies during dorsal-ventral axis formation. The replacement allowed vertebrate Nedd4L to acquire novel functions as a ubiquitin ligase of vertebrate-specific Smad proteins. Overall our data reveal that the architecture of the Dpp/BMP dorsal-ventral patterning network continued to evolve in the vertebrate lineage, after separation from flies, via the incorporation of new genes. © The Author 2014. Published by Oxford University Press on behalf of the Society for Molecular Biology and Evolution. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

Properties of thick welded joints on superheater collectors made from new generation high alloy martensitic creep-resisting steels for supercritical parameters

Energy Technology Data Exchange (ETDEWEB)

Dobrzanski, Janusz; Zielinski, Adam [Institute for Ferrous Metallurgy, Gliwice (Poland); Pasternak, Jerzy [Boiler Engineering Company RAFAKO S.A., Raciborz (Poland)

2010-07-01

The continuously developing power generation sector, including boilers with supercritical parameters, requires applications of new creep-resistant steel grades for construction of boilers steam superheater components. This paper presents selected information, experience within the field of research and implementation of a new group of creep-resistant as X10CrMoVNb9-1(P91), X10CrWMoVNb9-2(P92) and X12CrCoWVNb12-2-2(VM12) grades, containing 9-12%Cr. During welding and examination process the results of mechanical properties, requested level for base material and welded joints, as well as: tensile strength, impact strength and technological properties have been evaluated. Additional destructive examinations, with evaluation of structure stability, hardness distribution, for base material and welded joints after welding, heat treatment, again process have been determined. Recommendations due to the implementation influence of operating parameters of the main boiler components are part of this paper. (orig.)

Design of PFBR steam generator

International Nuclear Information System (INIS)

Chetal, S.C.; Bhoje, S.B.; Mitra, T.K.; Paranjpe, S.R.; Vaidyanathan, G.

1990-01-01

Vertical straight tube with an expansion bend in sodium path is the design selected for the steam generators of 500 MWe Prototype Fast Breeder Reactor (PFBR). There are 4 secondary loops with each loop consisting of 3 modules. With sodium reheat incorporated each module comprises of one evaporator, superheater and reheater. Material of construction is 2.25Cr-1Mo for evaporator and 9Cr-1Mo for superheater and reheater. The tube to tubesheet weld is internal bore butt weld with tubesheet having raised spigot. Aim is to have reliable design with higher plant availability. Design considerations leading to the choice of design features selected are discussed in the paper and a ''reference'' design has been described. (author). 2 figs, 1 tab

Relationship Between Timing of Multiple Retention Bonuses and the Quality of Officers Retained on the Cost Savings for the Navy

Science.gov (United States)

2016-12-01

psychologically anchor an item’s value at a high price before lowering it until one bidder accepts, earning its name from the way cut flowers are sold...Using the standard bonus system compared to the uniform price auction, the Quality Adjusted Discount (QUAD) auction, and Combinatorial Retention...NUMBER OF PAGES 85 16. PRICE CODE 17. SECURITY CLASSIFICATION OF REPORT Unclassified 18. SECURITY CLASSIFICATION OF THIS PAGE Unclassified 19

Experimental study of simulant melt stream-water thermal interaction in pool and narrow geometries

International Nuclear Information System (INIS)

Narayanan, K.S.; Jasmin Sudha, A.; Murthy, S.S.; Rao, E.H.V.M.; Lydia, G.; Das, S.K.; Harvey, J.; Kannan, S.E.

2005-01-01

Full text of publication follows: Small scale experiments were carried out to investigate the thermal interaction characteristics of a few kilograms of Sn Pb, Bi and Zn as hot melt, in the film boiling region of water in an attempt to simulate a coherent fuel coolant interaction during a postulated severe accident in a nuclear reactor. Melt stream solidification and detached debris generation were studied with different melt superheat up to 200 deg. C, at different coolant temperatures of 30 deg. C, 50 deg. C, 70 deg. C, 90 deg. C, in pool geometry and in long narrow coolant column. The material was heated in an Alumina crucible and poured through a hot stainless steel funnel with a nozzle diameter of 7.7 mm, into the coolant. A stainless steel plate was used to collect the solidified mass after the interaction. Temperature monitoring was done in the coolant column close to the melt stream. The melt stream movement inside the coolant was imaged using a video camera at 25 fps. Measured melt stream entry velocity was around 1.5 m/sec. For low melt superheat and low coolant temperature, solidified porous tree like structure extended from the collector plate up to the melt release point. For water temperature of 70 deg. C, the solidified bed height at the center was found to decrease with increase in the melt superheat up to 150 deg. C. Fragmentation was found to occur when the melt superheat exceeded 200 deg. C. Particle size distribution was obtained for the fragmented debris. In 1D geometry, with 50 deg. C superheat, columnar solidification was observed with no fine debris. The paper gives the details of the results obtained in the experiments and highlights the role of Rayleigh-Taylor, Kelvin-Helmholtz instabilities and the melt physical properties on the fragmentation kinetics. (authors)

Bone-targeted therapy for metastatic breast cancer—Where do we go from here? A commentary from the BONUS 8 meeting

Directory of Open Access Journals (Sweden)

Xiaofu Zhu

2014-03-01

Full Text Available The annual Bone and The Oncologist New Updates (BONUS 8 conference focuses on the current understanding and dilemmas in the treatment and prevention of bone metastasis in cancer, as well as novel research on bone homeostasis and cancer-induced bone loss. We present commentaries from experts for their own take on where they feel the field of bone-targeted therapies for metastatic breast cancer is moving, or needs to move, if we are to make further progress.

Oxalate Content of the Herb Good-King-Henry, Blitum Bonus-Henricus

Directory of Open Access Journals (Sweden)

Wanying Li

2015-05-01

Full Text Available The total, soluble and insoluble oxalate contents of the leaves, stems and buds of Good-King-Henry (Blitum Bonus-Henricus were extracted and measured using HPLC chromatography. The large, mature leaves contained 42% more total oxalate than in the small leaves and the soluble oxalate content of the large leaves was 33% higher than the smaller leaves. Cooking the mixed leaves, stems and buds in boiling water for two minutes significantly (p < 0.05 reduced the total oxalate when compared to the raw plant parts. Pesto sauce made from mixed leaves contained 257 mg total oxalate/100 g fresh weight; this was largely made up of insoluble oxalates (85% of the total oxalate content. Soup made from mixed leaves contained lower levels of total oxalates (44.26 ± 0.49 mg total oxalate/100 g fresh weight and insoluble oxalate made up 49% of the oxalate contents. The levels of oxalates in the Good-King-Henry leaves were high, suggesting that the leaves should be consumed occasionally as a delicacy because of their unique taste rather than as a significant part of the diet. However, the products made from Good-King-Henry leaves indicated that larger amounts could be consumed as the oxalate levels were reduced by dilution and processing.

A process for superheating steam in a nuclear power station circuit and device for putting in operation this process

International Nuclear Information System (INIS)

Monteil, Marcel; Forestier, Jean; Leblanc, Bernard; Monteil, Pierre

1975-01-01

A process is described for superheating steam in a nuclear power station circuit, comprising a turbine with a high pressure chamber and a low pressure chamber. It consists in superheating the steam between the high and low pressure chambers of the turbine by using as heating fluid water under pressure at vaporisation temperature, directly taken from the recirculation or circulation flow water of the reactor or of the steam generators. The process is adapted to a pressurised water reactor using a once-through steam generator comprising in succession an economiser, a vaporiser and a superheater, the superheating water being taken at the vaporiser intake. It is also adapted for a boiling water reactor, in that the water is taken directly from the reactor vessel and at a suitable level in the recirculation water [fr

Development of the heavy-water organic-cooled reactor. Status report from the United States of America

Energy Technology Data Exchange (ETDEWEB)

Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

1967-01-01

loops required. A second benefit of the pressure tube concept is that it is easily adaptable to the use of on-power refuelling, with the consequent potential of achieving high plant availability and optimizing the fuel management programme for maximum neutron economy. Many studies have been carried out in an attempt to optimize the selection of the coolant for a heavy-water-moderated reactor. The choice of an organic coolant for the US heavy-water reactor development programme offers several advantages. By limiting the heavy water to its function as moderator, its containment is required only at low temperatures and pressures, thus minimizing both inventory and losses of this expensive material. The compatibility of the organic coolant with uranium, plutonium, and thorium metals, and with their oxides and carbides provides for maximum flexibility in the selection of fuel material for this reactor. The HWOCR can therefore take full advantage of whichever fuel cycle in the long run demonstrates the most favourable economics. The low vapour pressure of the organic coolant and its compatibility with standard materials of construction provide for the design of low pressure primary coolant loops using carbon and low alloy steels while obtaining the thermodynamic efficiency of a superheat steam cycle. The low level of induced radioactivity normally present in the organic coolant permits normal contact maintenance of the primary coolant loops while the plant is in operation.

Corrosion evaluation of heat recovery steam generator superheater tube in two methods of testing: Tafel polarization and electrochemical impedance spectroscopy (EIS)

Science.gov (United States)

Santoso, Rio Pudjidarma; Riastuti, Rini

2018-05-01

The purpose of this research is to evaluate the corrosion process which occurs on the water side of Heat Recovery Steam Generator (HRSG) superheater tube. The tube was 13CrMo44 and divided into 3 types of specimen: new tube, used tube (with oxide layer on surface), cleaned-used tube (without oxide layer on surface). The evaluation of corrosion parameters wasperformed using deaerated ultra-high purity water (boiler feed water) in two methods of testing: Tafel polarization and Electrochemical Impedance Spectroscopy (EIS). Tafel polarization was excellent as its capability to show the value of corrosion current and the corrosion rate explicitly, on the other hand, EIS was excellent as its capability to explain for corrosion mechanism on metal interface in detail. Both methods showed that the increase of electrolyte temperature from 25°C to 55°C would increase the corrosion rate with the mechanism of decreasing polarization resistance due to thinning out the passive film thickness and enlarge the area of reduction reaction of cathode. Magnetite oxide scale which is laid on the surface of used tube specimen shows protective nature to reduce the corrosion rate, and clear up this oxide would increase the corrosion rate back as new tube.

Application of an empirical model in CFD simulations to predict the local high temperature corrosion potential in biomass fired boilers

International Nuclear Information System (INIS)

Gruber, Thomas; Scharler, Robert; Obernberger, Ingwald

2015-01-01

To gain reliable data for the development of an empirical model for the prediction of the local high temperature corrosion potential in biomass fired boilers, online corrosion probe measurements have been carried out. The measurements have been performed in a specially designed fixed bed/drop tube reactor in order to simulate a superheater boiler tube under well-controlled conditions. The investigated boiler steel 13CrMo4-5 is commonly used as steel for superheater tube bundles in biomass fired boilers. Within the test runs the flue gas temperature at the corrosion probe has been varied between 625 °C and 880 °C, while the steel temperature has been varied between 450 °C and 550 °C to simulate typical current and future live steam temperatures of biomass fired steam boilers. To investigate the dependence on the flue gas velocity, variations from 2 m·s −1 to 8 m·s −1 have been considered. The empirical model developed fits the measured data sufficiently well. Therefore, the model has been applied within a Computational Fluid Dynamics (CFD) simulation of flue gas flow and heat transfer to estimate the local corrosion potential of a wood chips fired 38 MW steam boiler. Additionally to the actual state analysis two further simulations have been carried out to investigate the influence of enhanced steam temperatures and a change of the flow direction of the final superheater tube bundle from parallel to counter-flow on the local corrosion potential. - Highlights: • Online corrosion probe measurements in a fixed bed/drop tube reactor. • Development of an empirical corrosion model. • Application of the model in a CFD simulation of flow and heat transfer. • Variation of boundary conditions and their effects on the corrosion potential

On the influence of water subcooling and melt jet parameters on debris formation

Energy Technology Data Exchange (ETDEWEB)

Manickam, Louis, E-mail: louis@safety.sci.kth.se; Kudinov, Pavel; Ma, Weimin; Bechta, Sevostian; Grishchenko, Dmitry

2016-12-01

Highlights: • Melt and water configuration effects on debris formation is studied experimentally. • Melt superheat and water subcooling are most influential compared to jet size. • Melt-water configuration and material properties influence particle fracture rate. • Results are compared with large scale experiments to study effect of spatial scales. - Abstract: Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-Jet facility by discharging binary-oxide mixtures of WO{sub 3}–Bi{sub 2}O{sub 3} and WO{sub 3}–ZrO{sub 2} into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased subcooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

CLUMPED LIGHT WATER MODERATED UO$sub 2$ SUPERHEAT CRITICALS. PART I. EXPERIMENTS

Energy Technology Data Exchange (ETDEWEB)

Warzek, F. G.; Johnston, H. F.

1963-11-15

The following critical and subcritical measurements were made in the EVESR core: reactivity with no control rods; full core reactivity with control rods; and power distribution in the full core with control rods. The fuel was UO/ sub 2/, and the elements were of the superheating type. The reactor was light- water-cooled and -moderated. (T.F.H.)

Functional performance of the helical coil steam generator, Consolidated Nuclear Steam Generator (CNSG) IV system. Executive summary report

International Nuclear Information System (INIS)

Watson, G.B.

1975-10-01

The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes and the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics

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

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Investigation of effective factors of transient thermal stress of the MONJU-System components

Energy Technology Data Exchange (ETDEWEB)

Inoue, Masaaki; Hirayama, Hiroshi; Kimura, Kimitaka; Jinbo, M. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

1999-03-01

Transient thermal stress of each system Component in the fast breeder reactor is an uncertain factor on it's structural design. The temperature distribution in a system component changes over a wide range in time and in space. An unified evaluation technique of thermal, hydraulic, and structural analysis, in which includes thermal striping, temperature stratification, transient thermal stress and the integrity of the system components, is required for the optimum design of tho fast reactor plant. Thermal boundary conditions should be set up by both the transient thermal stress analysis and the structural integrity evaluation of each system component. The reasonable thermal boundary conditions for the design of the MONJU and a demonstration fast reactor, are investigated. The temperature distribution analysis models and the thermal boundary conditions on the Y-piece structural parts of each system component, such as reactor vessel, intermediate heat exchanger, primary main circulation pump, steam generator, superheater and upper structure of reactor core, are illustrated in the report. (M. Suetake)

Jeopardy not bonus status for African American women in the work force: why does the myth of advantage persist?

Science.gov (United States)

Sanchez-Hucles, J V

1997-10-01

African American women in the United States have a long history of employment outside of their homes. Their experiences are unique from other groups of majority and minority men and women due to the interaction of race, gender, and class. Despite long-standing and continuing struggles against discrimination, harassment, low pay, tokenism, and stereotypes, a myth that African American women enjoy a bonus or advantaged status in the work force has developed and persisted. In this article, Black women's work force experiences are examined from a social constructionist framework, misperceptions of Black women are critiqued, explanations are developed that explain the unique status of African American women and recommendations are proposed to eradicate the discrimination and marginal status that Black women have endured in the work force.

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Power plant and system for accelerating a cross compound turbine in such plant, especially one having an HTGR steam supply

International Nuclear Information System (INIS)

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

1977-01-01

An electric power plant having a cross compound steam turbine and a steam source that includes a high temperature gas-cooled nuclear reactor is described. The steam turbine includes high and intermediate-pressure portions which drive a first generating means, and a low-pressure portion which drives a second generating means. The steam source supplies superheat steam to the high-pressure turbine portion, and an associated bypass permits the superheat steam to flow from the source to the exhaust of the high-pressure portion. The intermediate and low-pressure portions use reheat steam; an associated bypass permits reheat steam to flow from the source to the low-pressure exhaust. An auxiliary turbine driven by steam exhausted from the high-pressure portion and its bypass drives a gas blower to propel the coolant gas through the reactor. While the bypass flow of reheat steam is varied to maintain an elevated pressure of reheat steam upon its discharge from the source, both the first and second generating means and their associated turbines are accelerated initially by admitting steam to the intermediate and low-pressure portions. The electrical speed of the second generating means is equalized with that of the first generating means, whereupon the generating means are connected and acceleration proceeds under control of the flow through the high-pressure portion. 29 claims, 2 figures

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

A display to support knowledge based behavior

International Nuclear Information System (INIS)

Lindsay, R.W.

1990-01-01

A computerized display has been created for the Experimental Breeder Reactor II (EBR-II) that incorporates information from plant sensors in a thermodynamic model display. The display is designed to provide an operator with an overall view of the plant process as a heat engine. The thermodynamics of the plant are depicted through the use of ionic figures, animated by plant signals, that are related to the major plant components and systems such as the reactor, intermediate heat exchanger, secondary system, evaporators, superheaters, steam system, steam drum, and turbine-generator. This display supports knowledge based reasoning for the operator as well as providing the traditional rule and skill based behavior, and includes side benefits such a inherent signal validation

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

High temperature corrosion investigations at AW2-bio. Final report; Biomass boiler

Energy Technology Data Exchange (ETDEWEB)

Borg, U.

2011-01-15

The measured corrosion rates in the test superheaters and ordinary superheaters of Avedoere 2 biomass boiler reveal that the corrosion rate increases with metal temperature and is significantly accelerated above steam temperatures of 540 deg. C. For the boiler with a live steam temperature of 540 deg. C, the measured corrosion rates in superheater 2 and 3 were up to 1mm pr. 10000 hours. It was observed that the flue gas temperature and heat flux had a significant effect on the corrosion rates through the surface metal temperature. Thus, the highest corrosion rates in the ordinary superheaters were not found at the position of the highest steam temperature in the outlet of superheater 3, but at the outlet of superheater 2. A steam temperature of approximately 580 deg. C at the outlet of one of the test superheater loops caused a tube fracture after a few months. A HVOF coating was applied to a section of superheater 2 and at a higher temperature in the test superheater loop. Analyses of the tube section after exposure showed that parts of the coating were not present and corrosion of the underlying TP347H FG was apparent. This indicates that the coating had spalled during operation. Furthermore, chlorine diffusion through the coating was observed causing attack at the coating-alloy interface. The project work has shown that it is not possible to increase the live steam temperature of the biomass fired boiler to more than 540 deg. C without a significant increase in superheater corrosion rates for the applied tube materials and coatings. (Author)

Detection and Repair of Ligament Cracks in a 109mm Thick Superheater Outlet Header

International Nuclear Information System (INIS)

Day, Peter

2006-01-01

Conventional thermal power station boilers are constructed of drums and a series of headers which are interconnected with many hundreds of tubes. Typically feed water enters the boiler at about 250 deg C at a pressure of around 250 bar with steam outlet temperatures of 540 deg C and a pressure of 170 bar. Superheater outlet headers may be subjected to quite arduous conditions during service. Not only are they exposed to high pressure stresses but also to high thermal stresses due to varying thermal gradients through the section thickness particularly at start up and during two shift operation. The area that is exposed to the greatest thermal gradients is the narrow ligament that exists between the tube hole penetrations in the header bore. In the mid the 1980's industry wide surveys found cracking in a large percentage (25-50%) of headers after 15 years of service. Detection and sizing of ligament cracking and estimates of the rate of growth are therefore a major consideration especially in plant that is two shifted. In order to manage the risk both remote visual and ultrasonic inspection are performed during each major unit overhaul. Conclusion: Ultrasonic techniques used for this inspection need to be carefully evaluated with respect to their effectiveness. Conventional pulse echo is capable of detection but using for example a technique such as AS2207 level 1 will not show the defect size. Time of flight diffraction has shown itself to be effective in accurately sizing ligament cracking. However the complex geometry of header ligaments appears to cause a narrowing of the beam with the effect that crack tip responses can be concentrated at the centre of the ligament. Therefore great care needs to be taken during data interrogation because errors in sizing can occur. Wherever possible both 'B' and 'D' scan data should be collected. It appears that the greatest accuracy is obtained with respect to defect growth from the B scan image. With respect to the welding a

The desorption of caesium from Peach Bottom HTGR steam generator materials

International Nuclear Information System (INIS)

Clark, M.J.

1979-03-01

The work at Harwell on the Peach Bottom End-of-Life Program in co-operation with the General Atomic Company (U.S.A.) is described. Materials taken from the Economiser, Evaporator and Superheater Sections of the Peach Bottom Unit No. 1. High Temperature Gas Cooled Reactor (HTGR) Heat Exchanger were placed in a reducing atmosphere comparable to the composition of an HTGR helium coolant gas, and the desorption of caesium isotopes measured under known conditions of flow, temperature and oxygen pressure. (author)

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

A worked example using the SP249 advanced assessment route: the carregado unit 6 final superheater outlet header

Energy Technology Data Exchange (ETDEWEB)

Brear, J.M.; Jarvis, P.; Jones, G.T. [ERA Technology (United Kingdom); Jovanovic, A.S.; Friemann, M.; Kluttig, B.; Ober, M. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Batista, A. [EDP-PROET (Portugal); Araujo, C.L. de; Pires, A. [ISQ (Portugal)

1995-12-31

As a key part of its information resource, the SP249 Project contains a number of case studies, drawn from the collective experience of the partners and from the literature. The user of the system may search this data-base by component type and material or by assessment method, to find a practical example close to his own current problem. He can thus draw upon past experience as well as state-of-the-art knowledge to obtain advice. To facilitate this, a set of key-words has been defined to create links between the case studies and the overall assessment methodology. These relate to damage and failure types and causes as well as to techniques of investigation and assessment. For demonstration, validation and didactic purposes, certain of these case studies - one per end-user utility in the project - have been chosen for full elaboration as `worked-examples`. These real component evaluations are worked through by an expert group from the project team so as to provide the utility staff with `hands-on` training in both the practical techniques of component life. The assessment and the use of the knowledge based system. The exercise also provides valuable opportunity for feedback, allowing refinement of the technology package and the software. Amongst these worked examples, an assessment of EDP`s Carregado Unit 6 Final Superheater Outlet Header has been chosen for special attention - as the operators have kindly allowed direct CSS to the component during two outages. This article summarises the Carregado Case Study. It is intended to serve as a demonstration and as to how the Advanced Assessment Route (AAR) is used in practice. The actions performed and results obtained are summarised

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

SASSYS LMFBR systems analysis code

International Nuclear Information System (INIS)

Dunn, F.E.; Prohammer, F.G.

1982-01-01

The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

A display to support knowledge based behavior

International Nuclear Information System (INIS)

Lindsay, R.W.

1990-01-01

This paper reports on a computerized display that has been created for the Experimental Breeder Reactor II that incorporates information from plant sensors in a thermodynamic model display. The display is designed to provide an operator with an overall view of the plant process as a heat engine. The thermodynamics of the plant are depicted through the use of iconic figures, animated by plant signals, that are related to the major plant components and systems such as the reactor, intermediate heat exchanger, secondary system, evaporators, superheaters, steam system, steam drum, and turbine-generator. This display supports knowledge based reasoning for the operator as well as providing data for the traditional rule and skill based behavior, and includes side benefits such as inherent signal validation

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

International Nuclear Information System (INIS)

Anon.

1986-01-01

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

LMFBR steam generators in the United Kingdom

International Nuclear Information System (INIS)

Anderson, R.; Hayden, O.

2002-01-01

Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Criterio no asintótico para el diseño de sistemas Bonus-Malus mediante GPBM

Directory of Open Access Journals (Sweden)

Antonio Heras Martínez

2002-01-01

Full Text Available El diseño de sistemas Bonus-Malus habitualmente está basado en la distribución estacionaria de la probabilidad. Se supone que una póliza ha alcanzado el estado estacionario (n0 cuando la probabilidad de que dicha póliza pertenezca a la clase Ci es constante para todo n mayor o igual que n0 . Esta distribución estacionaria de probabilidad es sencilla de calcular. Su cálculo está basado en la Teoría de Cadenas de Markov y requiere que el sistema cumpla ciertas condiciones (sea Cadena de Markov irreducible y todos sus estados sean recurrentes en tiempo esperado finito. Podría ocurrir que estas condiciones no se cumplan y en este caso no se podría utilizar la distribución estacionaria. Por otra parte, los criterios basados en la distribución estacionaria no tienen en cuenta a las pólizas nuevas en la empresa ni a las que llevan poco tiempo. Por estos motivos, se propone un nuevo criterio que tenga en cuenta la edad de las pólizas de la cartera.

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Nuclear and thermal power plant power ramping capability

International Nuclear Information System (INIS)

Golovach, E.A.

1983-01-01

The possibilities of step power increase by NPP and TPP units under emergency conditions of power grids operation are considered. The data analysis has shown that power units ramping capability with WWER-440, WWER-1000 and RBMK-1000 reactors is higher than that of 300 MW power units on fossil fuel, at the initial time interval (0-30 s). These NPP power units satisfy as to ramping capability the energy system requirements. Higher NPP power units ramping capability is explained by the fact that relative pressure before turbine valves is decreased less than in straight-through boilers while the steam volumes time constant of steam separator-superheaters is less than that of intermediate superheatings. Higher power unit ramping capability with WWER-440 and RBMK-1000 reactors as compared with the WWER-1000 reactor is pointed out as well as the increase of WWER-1000 power unit capability using high-speed turbines

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Inception and development of voids in flashing liquids

International Nuclear Information System (INIS)

Jones, O.C. Jr.

1979-06-01

Recent work aimed at correctly describing nonequilibrium vapor generation rates in flashing liquids in decompressing flows similar to those which might be encountered in a loss of coolant accident in a nuclear reactor is summarized. Analysis is reviewed which describes the flashing inception superheat in terms of the turbulence intensity for a given expansion rate and initial temperature, and interfacial area density and interfacial heat flux, and the volumetric vapor generation rates. Comparisons with existing data are included and further experiments being undertaken are described, including typical recent results

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

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

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

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

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Exergoeconomic assessment and parametric study of a Gas Turbine-Modular Helium Reactor combined with two Organic Rankine Cycles

International Nuclear Information System (INIS)

Mohammadkhani, F.; Shokati, N.; Mahmoudi, S.M.S.; Yari, M.; Rosen, M.A.

2014-01-01

An exergoeconomic analysis is reported for a combined system with a net electrical output of 299 MW in which waste heat from a Gas Turbine-Modular Helium Reactor (GT-MHR) is utilized by two Organic Rankine Cycles (ORCs). A parametric study is also done to reveal the effects on the exergoeconomic performance of the combined system of such significant parameters as compressor pressure ratio, turbine inlet temperature, temperatures of evaporators, pinch point temperature difference in the evaporators and degree of superheat at the ORC (Organic Rankine Cycle) turbines inlet. Finally the combined cycle performance is optimized from the viewpoint of exergoeconomics. The results show that the precooler, the intercooler and the ORC condensers exhibit the worst exergoeconomic performance. For the overall system, the exergoeconomic factor, the capital cost rate and the exergy destruction cost rate are determined to be 37.95%, 6876 $/h and 11,242 $/h, respectively. Also, it is observed that the unit cost of electricity produced by the GT-MHR turbine increases with increasing GT-MHR turbine inlet temperature but decreases as the other above mentioned parameters increase. - Highlights: • An exergoeconomic analysis is performed for the GT-MHR/ORC (Organic Rankine Cycle) combined cycle. • The effects of decision parameters on the exergoeconomic performance are studied. • The highest exergy destructions occur in the precooler, intercooler and condenser. • Superheating the working fluid at the ORC turbine inlet is not necessary. • Thermodynamic and exergoeconomic optimal conditions differ from each other

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Power technology complex for production of motor fuel from brown coals with power supply from NPPs

International Nuclear Information System (INIS)

Troyanov, M.F.; Poplavskij, V.M.; Sidorov, G.I.; Bondarenko, A.V.; Chebeskov, A.N.; Chushkin, V.N.; Karabash, A.A.; Krichko, A.A.; Maloletnev, A.S.

1998-01-01

With the present-day challenge of efficient use of low-grade coals and current restructuring of coal industry in the Russian Federation, it is urgent to organise the motor fuel production by the synthesis from low grade coals and heavy petroleum residues. With this objective in view, the Institute of Physics and Power Engineering of RF Minatom and Combustible Resources Institute of RF Mintopenergo proposed a project of a standard nuclear power technology complex for synthetic liquid fuel (SLF) production using fast neutron reactors for power supply. The proposed project has two main objectives: (1) Engineering and economical optimization of the nuclear power supply for SLF production; and (2) Engineering and economical optimization of the SLF production by hydrogenisation of brown coals and heavy petroleum residues with a complex development of advanced coal chemistry. As a first approach, a scheme is proposed with the use of existing reactor cooling equipment, in particular, steam generators of BN-600, limiting the effect on safety of reactor facility operation at minimum in case of deviations and abnormalities in the operation of technological complex. The possibility to exclude additional requirements to the equipment for nuclear facility cooling was also taken into account. It was proposed to use an intermediate steam-water circuit between the secondary circuit sodium and the coolant to heat the technological equipment. The only change required for the BN-600 equipment will be the replacement of sections of intermediate steam superheaters at the section of main steam superheaters. The economic aspects of synthetic motor fuel production proposed by the joint project depend on the evaluation of integral balances: thermal power engineering, chemical technology, the development of advanced large scale coal chemistry of high profitability; utilisation of ash and precious microelements in waste-free technology; production of valuable isotopes; radical solution of

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

An experimental observation of the effect of flow direction for evaporation heat transfer in plate heat exchanger

International Nuclear Information System (INIS)

Lin, Yueh-Hung; Li, Guang-Cheng; Yang, Chien-Yuh

2015-01-01

This study provides an Infrared Thermal Image observation on the evaporation heat transfer of refrigerant R-410A in plate heat exchanger with various flow arrangement and exit superheat conditions. An experimental method was derived for estimating the superheat region area of two-phase refrigerant evaporation in plate heat exchanger. The experimental results show that the superheat region area for parallel flow is much larger than that for counter flow as that estimated by Yang et al. [9]. There is an early superheated region at the central part of the plate heat exchanger for parallel flow arrangement. This effect is not significant for counter flow arrangement. The Yang et al. [9] method under estimated the superheat area approximately 40%–53% at various flow rates and degree of exit superheat. Even though the flow inside a plate heat exchanger is extremely turbulent because of the chevron flow passages, the assumption of uniform temperature distribution in the cross section normal to the bulk flow direction will cause significant uncertainties for estimating the superheat area for refrigerant evaporating in a plate heat exchanger

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

International Nuclear Information System (INIS)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

1999-01-01

This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

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

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

System of operative computer control of power distribution fields in the Beloyarsk nuclear power plant

International Nuclear Information System (INIS)

Kulikov, N.Ya.; Snitko, Eh.I.; Rasputnis, A.M.; Solodov, V.P.

1976-01-01

Describes the system of intrareactor control over the reactors of the Byeloyarskaya Atomic Station. In the second block of the station, use is made of direct charge emission detectors installed in the central apertures of the superheater channels and operating reliably at temperatures up to 750 deg C. The detectors of the first and the second block are connected to the computer which sends the results of processing the signals to the printer, while the signals for deviations go to the mnemonic tablaux of the reactors. The good working order of the detectors is checked by comparison with zero as well as with the mean detector current for the reactor concerned. The application of the intrareactor control system has allowed the stable thermal power to be increased from 480-500 to 530 Mw and makes it possible to control and maintain the neutron field formed with a relative error of 3-4%. The structural scheme of the system of intrareactor control is given

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

RA Reactor

International Nuclear Information System (INIS)

1978-02-01

In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

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%

Integral analysis of debris material and heat transport in reactor vessel lower plenum

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1994-01-01

An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. ((orig.))

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

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

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

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

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Numerical analyses of flashing jet structure and droplet size characteristics

International Nuclear Information System (INIS)

Duan Riqiang; Jiang Shengyao; Koshizuka, Seiichi; Oka, Yoshiaki; Yamaguchi, Akira; Takata, Takashi

2006-01-01

In this paper, flashing jets are numerically simulated using the MPS method. The boiling mode for flashing is identified as surface boiling mode, based on the postulation of jets from a short nozzle under high depressurization. The Homogeneous Non-equilibrium Relaxation Model (HRM) is used for calculating the evaporation rate of flashing. The numerical simulation results show that flashing jets comprise an inner intact core which is surrounded by two-phase droplet flow. The effect of degree of superheat on the jet topological geometry is investigated. With increasing degree of superheat, the topological shape of flashing jets evolves from cylindrical core for low degree of superheat to cone-shaped core for high degree of superheat, and meanwhile the extinction length comes to decrease and tends asymptotically constant as the injection temperature approaches the saturation temperature corresponding to the injection pressure. The analyses of the droplet size distribution engendered from primary breakup of flashing jets show that: two peaks exist for droplet size distribution at lower degree of superheat; however, merely one peak for higher degree of superheat. From droplet size distribution, it is revealed that the primary breakup mechanism of flashing jets can be attributed to dominant mechanical breakup mode plus enhancement via surface evaporation. (author)

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1983-01-01

Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Heterogeneous reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

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

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

Reactor container

International Nuclear Information System (INIS)

Fukazawa, Masanori.

1991-01-01

A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

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

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

New about research reactors

International Nuclear Information System (INIS)

Egorenkov, P.M.

2001-01-01

The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Design guide for Category III reactors: pool type reactors

International Nuclear Information System (INIS)

Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

1978-11-01

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

Reactor safety method

International Nuclear Information System (INIS)

Vachon, L.J.

1980-01-01

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

Experimental study of nonequilibrium post-chf heat transfer in rod bundles

International Nuclear Information System (INIS)

Unal, C.; Tuzla, K.; Badr, O.; Neti, S.; Chen, J.

1986-01-01

Verifications and improvements of nonequilibrium heat transfer models, for post-critical-heat-flux convective boiling, has been greatly affected by the lack of experimental data regarding the degree of thermodynamic nonequilibrium. Recent studies had been successful in measuring vapor superheats in a vertical single tube. This paper extends the nonequilibrium convective boiling data to a rod bundle geometry. Vapor superheat measurements were obtained in a rod bundle with nine heated rods and a heated shroud. Tests were carried out with water at low mass fluxes with a wide range of dryout conditions. Significant nonequilibrium was observed, with vapor superheats of up to 600 0 C. Parametric effects of mass flux, heat flux and inlet conditions on vapor superheat are presented

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

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

Analysis of dynamic stability and safety of reactor system by reactor simulator

International Nuclear Information System (INIS)

Raisic, N.

1963-11-01

In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Atomic reactor thermal engineering

International Nuclear Information System (INIS)

Kim, Gwang Ryong

1983-02-01

This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

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.)

Reactor container

International Nuclear Information System (INIS)

Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

1997-01-01

In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

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)

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)

Subcooled boiling heat transfer and dryout on a constant temperature microheater

International Nuclear Information System (INIS)

Chen Tailian; Klausner, James F.; Chung, Jacob N.

2004-01-01

An experimental study of single-bubble subcooled boiling heat transfer (ΔT sub =31.5 K) on a small heater with constant wall temperature has been performed to better understand the boiling heat transfer associated with this unique configuration. The heater of 0.27 mm x 0.27 mm is set at different superheats to generate vapor bubbles on the microheater surface. For each superheat, the heater temperature is maintained constant by an electronic feedback control circuit while its power dissipation is measured at a frequency of 4.5 kHz. The single-bubble boiling is characterized by a transient bubble nucleation-departure period and a slow growth period. For the superheat range of 34-114 K in this study, at wall superheats below 84 K, the heater remains partially wetted following bubble departure and subsequent nucleation, and this period is characterized by a heat flux spike. At wall superheats above 90 K, the heater is blanketed with vapor following bubble departure and the heat flux experiences a dip during this period. At all superheats, the slow growth period is characterized by an almost uniform heat flux, and it has been observed that the heater surface is mostly covered by vapor. The unique heat transfer processes associated with boiling on this microheater are considerably different than those typically observed during boiling on a large heater

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Batheja, P.; Huber, R.; Rau, P.

1985-01-01

Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

Energy Technology Data Exchange (ETDEWEB)

Kliem, S.

2017-09-15

The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

The fast breeder reactor

International Nuclear Information System (INIS)

Collier, J.

1990-01-01

The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Heat transfer characteristics evaluation of heat exchangers of mock-up test facility with full-scale reaction tube for HTTR hydrogen production system (Contract research)

International Nuclear Information System (INIS)

Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro; Maeda, Yukimasa; Sato, Hiroyuki; Inagaki, Yoshiyuki; Hanawa, Hiromi; Fujisaki, Katsuo; Yonekawa, Hideo

2005-06-01

Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001 and hydrogen of 120 Nm 3 /h was successfully produced in overall performance test. Totally 7 times operational tests were performed from March 2002 to December 2004. A lot of operational test data on heat exchanges were obtained in these tests. In this report specifications and structures of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Heat transfer correlation equations for inside and outside tube were chosen from references. Spreadsheet programs were newly made to evaluate heat transfer characteristics from measured test data such as inlet and outlet temperature pressure and flow-rate. Overall heat-transfer coefficients obtained from the experimental data were compared and evaluated with the calculated values with heat transfer correlation equation. As a result, actual measurement values of all heat exchangers gave close agreement with the calculated values with correlation equations. Thermal efficiencies of the heat exchangers were adequate as they were well accorded with design value. (author)

Ulysse, mentor reactor

International Nuclear Information System (INIS)

Bouquin, B.; Rio, I.; Safieh, J.

1997-01-01

On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

Analysis of a high pressure ATWS [anticipated transient without scram] with very low make-up flow

International Nuclear Information System (INIS)

Wagner, K.C.

1988-10-01

A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

Science.gov (United States)

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

The fast reactor

International Nuclear Information System (INIS)

1980-02-01

The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Masood, Z.

2016-01-01

The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

Fast Spectrum Reactors

CERN Document Server

Todd, Donald; Tsvetkov, Pavel

2012-01-01

Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

Droplet condensation in rapidly decaying pressure fields

International Nuclear Information System (INIS)

Peterson, P.F.; Bai, R.Y.; Schrock, V.E.; Hijikata, K.

1992-01-01

Certain promising schemes for cooling inertial confinement fusion reactors call for highly transient condensation in a rapidly decaying pressure field. After an initial period of condensation on a subcooled droplet, undesirable evaporation begins to occur. Recirculation within the droplet strongly impacts the character of this condensation-evaporation cycle, particularly when the recirculation time constant is of the order of the pressure decay time constant. Recirculation can augment the heat transfer, delay the onset of evaporation, and increase the maximum superheat inside the drop by as much as an order of magnitude. This numerical investigation identifies the most important parameters and physics characterizing transient, high heat flux droplet condensation. The results can be applied to conceptual designs of inertial confinement fusion reactors, where initial temperature differences on the order of 1,500 K decay to zero over time spans the order of tens of milliseconds

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

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

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Refuelling nuclear reactors

International Nuclear Information System (INIS)

Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

1981-01-01

An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

Compilation of reactor physics data of the year 1984, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-12-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Compilation of reactor physics data of the year 1983, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-06-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

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

1992-01-01

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

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Fission reactors and materials

International Nuclear Information System (INIS)

Frost, B.R.T.

1981-12-01

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

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

Method of operating a reactor

International Nuclear Information System (INIS)

Oosumi, Katsumi; Yamamoto, Michiyoshi.

1980-01-01

Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

Reactor container

International Nuclear Information System (INIS)

Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

1987-01-01

Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

Department of reactor technology

International Nuclear Information System (INIS)

1980-01-01

The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

1964-02-15

Numerous research, central station power, propulsion, isotope production, and test reactor designs have been investigated in Argonne's zero-power reactor facilities, and related exponential and clean critical assemblies have provided basic data. To present a representative account of recent experiments and to demonstrate the wide variety of reactor design information obtainable in low flux systems, the following experimental programmes are reviewed: 1. A study of the properties of thoria-urania fuel in heavy water, with particular attention to the requirements for design of a second core for Argonne's Experimental Boiling Water Reactor; 2. A mock-up of a proposed high flux research reactor to confirm the design calculations, optimize the geometry and estimate the effect of fuel burn-up; 3. A determination of the power distribution patterns and reactivity effect of fuel element flooding for a combined boiling-superheat reactor test; 4. The design of a sodium cooled. U{sup 235} fueled, plutonium producing fast breeder reactor core as a first loading for Argonne's Experimental Breeder Reactor II; and 5. An investigation of the characteristics of a reactor with interacting thermal and fast neutron zones. In the discussion of these programmes, the circumstances which influenced the choice among exponentials, clean criticals, zero-power mock-ups and in situ experiments for the acquisition of the required data are explained, as is the role played by supporting analytical effort. The extent to which reactor design data can be attained before actual operation at power is illustrated by specific examples. Such data include shutdown margin, excess reactivity for operational requirements, temperature coefficients, control and safety rods' effectiveness, reactor kinetics, power production patterns, requirements for start-up source and instrument sensitivity, shielding needs and neutron economy. This review of recent activities in zero-power experimentation reveals the strong

Failure evaluation on a high-strength alloy SA213-T91 super heater tube of a power generation

Energy Technology Data Exchange (ETDEWEB)

Ahmad, J.; Purbolaksono, J.; Beng, L.C.; Ahmad, A. [University of Tenaga Nas, Kajang (Malaysia). Dept. of Mechanical Engineering

2010-07-01

This article presents failure investigation on a high-strength alloy SA213-T91 superheater tube. This failure is the first occurrence involving the material in Kapar Power Station Malaysia. The investigation includes visual inspections, hardness measurements, and microscopic examinations. The failed super-heater tube shows a wide open rupture with thin and blunt edges. Hardness readings on all the as-received tubes are used for estimating the operating metal temperature of the super-heater tubes. Microstructures of the failed tube show numerous creep cavities consisting of individual pores and chain of pores which form micro-and macro-cracks. The findings confirmed that the super-heater tube is failed by short-term overheating. Higher temperatures of the flue gas due to the inconsistent feeding of pulverized fuels into the burner is identified to cause overheating of the failed tube.

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

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.

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

One piece reactor removal

International Nuclear Information System (INIS)

Chia, Wei-Min; Wang, Song-Feng

1993-01-01

The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

Nuclear reactors

International Nuclear Information System (INIS)

Middleton, J.E.

1977-01-01

Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

Reactor decommissioning

International Nuclear Information System (INIS)

Lawton, H.

1984-01-01

A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

2000-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1998-01-01

Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr