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

International Nuclear Information System (INIS)

Wang, L.C.

1984-01-01

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

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

pH-oscillations in the bromate-sulfite reaction in semibatch and in gel-fed batch reactors

Science.gov (United States)

Poros, Eszter; Kurin-Csörgei, Krisztina; Szalai, István; Rábai, Gyula; Orbán, Miklós

2015-06-01

The simplest bromate oxidation based pH-oscillator, the two component BrO3--SO32- flow system was transformed to operate under semibatch and closed arrangements. The experimental preconditions of the pH-oscillations in semibatch configuration were predicted by model calculations. Using this information as guideline large amplitude (ΔpH˜3), long lasting (11-24 h) pH-oscillations accompanied with only a 20% increase of the volume in the reactor were measured when a mixture of Na2SO3 and H2SO4 was pumped into the solution of BrO3- with a very low rate. Batch-like pH-oscillations, similar in amplitude and period time appeared when the sulfite supply was substituted by its dissolution from a gel layer prepared previously in the reactor in presence of high concentration of Na2SO3. The dissolution vs time curve and the pH-oscillations in the semibatch and closed systems were successfully simulated. Due to the simplicity in composition and in experimental technique, the semibatch and batch-like BrO3--SO32- pH-oscillators may become superior to their CSTR (continuous flow stirred tank reactor) version in some present and future applications.

Average cross-sections for /n, p/ reactions on calcium in a fission-type reactor spectrum

International Nuclear Information System (INIS)

Bruggeman, A.; Maenhaut, W.; Hoste, J.

1974-01-01

The average cross-section in a fission-type reactor spectrum sigmasub(F) was experimentally determined for the reactions 42 Ca/n,p/ 42 K, 43 Ca/n,p/ 43 K and 44 Ca/n,p/ 44 K. Calcium carbonate samples and fast neutron flux monitors were irradiated with and without cadmium shielding in the Thetis reactor (Institute for Nuclear Sciences, Rijksuniversiteit Gent). The potassium activities induced in the calcium carbonate samples were separated and purified by tetraphenylborate precipitation, after which they were measured with a Ge/Li/-detector of calibrated detection efficiency. On the basis of sigmasub(F)=0.64 mb for the reaction 27 Al/n,α/ 24 Na, the average cross-sections were as follows: 42 Ca/n,p/ 42 K: 2.82+-0.07 mb; 43 Ca/n,p/ 43 K: 1.89+-0.05 mb; 44 Ca/n,p/ 44 K: 0.065+-0.003 mb. (T.G.)

Solution of the Lambda modes problem of a nuclear power reactor using an h–p finite element method

International Nuclear Information System (INIS)

Vidal-Ferrandiz, A.; Fayez, R.; Ginestar, D.; Verdú, G.

2014-01-01

Highlights: • An hp finite element method is proposed for the Lambda modes problem of a nuclear reactor. • Different strategies can be implemented for increasing the accuracy of the solutions. • 2D and 3D benchmarks have been studied obtaining accurate results. - Abstract: Lambda modes of a nuclear power reactor have interest in reactor physics since they have been used to develop modal methods and to study BWR reactor instabilities. An h–p-Adaptation finite element method has been implemented to compute the dominant modes the fundamental mode and the next subcritical modes of a nuclear reactor. The performance of this method has been studied in three benchmark problems, a homogeneous 2D reactor, the 2D BIBLIS reactor and the 3D IAEA reactor

Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

Directory of Open Access Journals (Sweden)

Carré Frank

2014-01-01

Full Text Available Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

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

pH-oscillations in the bromate–sulfite reaction in semibatch and in gel-fed batch reactors

Energy Technology Data Exchange (ETDEWEB)

Poros, Eszter; Kurin-Csörgei, Krisztina; Szalai, István; Orbán, Miklós, E-mail: orbanm@chem.elte.hu [Department of Analytical Chemistry, Institute of Chemistry, L. Eötvös University, P.O. Box 32, H-1518 Budapest 112 (Hungary); Rábai, Gyula [Institute of Physical Chemistry, University of Debrecen, P.O. Box 7, H-4010 Debrecen (Hungary)

2015-06-15

The simplest bromate oxidation based pH-oscillator, the two component BrO{sub 3}{sup −}–SO{sub 3}{sup 2–} flow system was transformed to operate under semibatch and closed arrangements. The experimental preconditions of the pH-oscillations in semibatch configuration were predicted by model calculations. Using this information as guideline large amplitude (ΔpH∼3), long lasting (11–24 h) pH-oscillations accompanied with only a 20% increase of the volume in the reactor were measured when a mixture of Na{sub 2}SO{sub 3} and H{sub 2}SO{sub 4} was pumped into the solution of BrO{sub 3}{sup −} with a very low rate. Batch-like pH-oscillations, similar in amplitude and period time appeared when the sulfite supply was substituted by its dissolution from a gel layer prepared previously in the reactor in presence of high concentration of Na{sub 2}SO{sub 3}. The dissolution vs time curve and the pH-oscillations in the semibatch and closed systems were successfully simulated. Due to the simplicity in composition and in experimental technique, the semibatch and batch-like BrO{sub 3}{sup −}–SO{sub 3}{sup 2–} pH-oscillators may become superior to their CSTR (continuous flow stirred tank reactor) version in some present and future applications.

V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

International Nuclear Information System (INIS)

Teuchert, E.; Hansen, U.; Haas, K.A.

1980-03-01

V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

International Nuclear Information System (INIS)

Flach, G.P.

1999-01-01

A regional groundwater flow model encompassing approximately 100 mi 2 surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department

Control program of the neutron four-circle-diffractometer P32 at the SILOE reactor/CEN Grenoble

International Nuclear Information System (INIS)

Guth, H.; Paulus, H.; Reimers, W.; Heger, G.

1983-09-01

The four-circle diffractometer P32 for elastic neutron scattering on single crystals was installed at the SILOE reactor/CEN Grenoble in 1981. The control program, presented here, is a new update of the former program versions used at the FR2 reactor/Kernforschungszentrum Karlsruhe. Important improvements concerning reliability and handling of the diffractometer are added. (orig.) [de

Control program of the neutron four-circle-diffractometer P110 at the ORPHEE reactor/CEN Saclay

International Nuclear Information System (INIS)

Guth, H.; Paulus, H.; Reimers, W.; Heger, G.

1984-05-01

The four-circle diffractometer P110 for elastic neutron scattering on single crystals was installed at the ORPHEE reactor/CEN Saclay in 1982. The control progam, presented here, is a new update of the former program versions used at the FR2 reactor. Important improvements concerning reliability and handling of the diffractometer are added. (orig./HP) [de

Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

Energy Technology Data Exchange (ETDEWEB)

Flach, G.P.

1999-02-24

A regional groundwater flow model encompassing approximately 100 mi{sup 2} surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department.

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

Science.gov (United States)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min-1, 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

Science.gov (United States)

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

Energy Technology Data Exchange (ETDEWEB)

Flach, G.P.

2000-02-11

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

International Nuclear Information System (INIS)

Flach, G.P.

2000-01-01

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

International Nuclear Information System (INIS)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-01-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min −1 , 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance. (paper)

Optimization of food waste hydrolysis in leach bed coupled with methanogenic reactor: effect of pH and bulking agent.

Science.gov (United States)

Xu, Su Yun; Lam, Hoi Pui; Karthikeyan, O Parthiba; Wong, Jonathan W C

2011-02-01

The effects of pH and bulking agents on hydrolysis/acidogenesis of food waste were studied using leach bed reactor (LBR) coupled with methanogenic up-flow anaerobic sludge blanket (UASB) reactor. The hydrolysis rate under regulated pH (6.0) was studied and compared with unregulated one during initial experiment. Then, the efficacies of five different bulking agents, i.e. plastic full particles, plastic hollow sphere, bottom ash, wood chip and saw dust were experimented under the regulated pH condition. Leachate recirculation with 50% water replacement was practiced throughout the experiment. Results proved that the daily leachate recirculation with pH control (6.0) accelerated the hydrolysis rate (59% higher volatile fatty acids) and methane production (up to 88%) compared to that of control without pH control. Furthermore, bottom ash improved the reactor alkalinity, which internally buffered the system that improved the methane production rate (0.182 l CH(4)/g VS(added)) than other bulking agents. Copyright © 2010 Elsevier Ltd. All rights reserved.

Biodegradation of p-nitrophenol using Arthrobacter chlorophenolicus A6 in a novel upflow packed bed reactor

International Nuclear Information System (INIS)

Sahoo, Naresh Kumar; Pakshirajan, Kannan; Ghosh, Pranab Kumar

2011-01-01

A novel packed bed reactor (PBR) was designed with cross flow aeration at multiple ports along the depth to improve the hydrodynamic conditions of the reactor, and the biodegradation efficiency of Arthrobacter chlorophenolicus A6 on p-nitrophenol (PNP) removal in PBR at different PNP loading rates were evaluated. The novel PBR was designed to improve the hydrodynamic features such as mixing time profile (t m95 ), oxygen mass transfer coefficient (k L a), and overall gas hold up capacity (ε G ) of the reactor. PNP concentration in the influent was varied between 600 and 1400 mg l -1 whereas the hydraulic retention time (HRT) in the reactor was varied between 18 and 7.5 h. Complete removal of PNP was achieved in the reactor up to a PNP loading rate of 2787 mg l -1 d -1 . More than 99.9% removal of PNP was achieved in the reactor for an influent concentration of 1400 mg l -1 and at 18 h HRT. In the present study, PNP was utilized as sole source of carbon and energy by A. chlorophenolicus A6. Furthermore, the bioreactor showed good compatibility in handling shock loading of PNP.

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%

Continuous control of pH value and chloride concentration in a water coolant of nuclear reactors

International Nuclear Information System (INIS)

Moskvin, L.N.; Krasnoperov, V.M.; Fokina, K.G.; Vilkov, N.Ya.

1975-01-01

Potentiometry method with the use of flowing cells with two identical electrodes is the simplest and most safe for continuous pH value and chloride control in nuclear reactor circulating circuits. The constant potential on the comparison electrode may be provided by supplying the analyzed solution to it through the ion resin filter of mixed operation. The pos--sibility of a continuous pH value monitoring in a flowing cell with two glass electrodes in parallel is considered. To monitor clorides a cell with two porous chlorine-silver electrodes positioned in series is used. The cells of the design described are shown to be workable in water simulating coolants for water-cooled reactors

Effects of plastic composite support and pH profiles on pullulan production in a biofilm reactor.

Science.gov (United States)

Cheng, Kuan-Chen; Demirci, Ali; Catchmark, Jeffrey M

2010-04-01

Pullulan is a linear homopolysaccharide which is composed of glucose units and often described as alpha-1, 6-linked maltotriose. The applications of pullulan range from usage as blood plasma substitutes to environmental pollution control agents. In this study, a biofilm reactor with plastic composite support (PCS) was evaluated for pullulan production using Aureobasidium pullulans. In test tube fermentations, PCS with soybean hulls, defatted soy bean flour, yeast extract, dried bovine red blood cells, and mineral salts was selected for biofilm reactor fermentation (due to its high nitrogen content, moderate nitrogen leaching rate, and high biomass attachment). Three pH profiles were later applied to evaluate their effects on pullulan production in a PCS biofilm reactor. The results demonstrated that when a constant pH at 5.0 was applied, the time course of pullulan production was advanced and the concentration of pullulan reached 32.9 g/L after 7-day cultivation, which is 1.8-fold higher than its respective suspension culture. The quality analysis demonstrated that the purity of produced pullulan was 95.8% and its viscosity was 2.4 centipoise. Fourier transform infrared spectroscopy spectra also supported the supposition that the produced exopolysaccharide was mostly pullulan. Overall, this study demonstrated that a biofilm reactor can be successfully implemented to enhance pullulan production and maintain its high purity.

Biodegradation of p-nitrophenol using Arthrobacter chlorophenolicus A6 in a novel upflow packed bed reactor.

Science.gov (United States)

Sahoo, Naresh Kumar; Pakshirajan, Kannan; Ghosh, Pranab Kumar

2011-06-15

A novel packed bed reactor (PBR) was designed with cross flow aeration at multiple ports along the depth to improve the hydrodynamic conditions of the reactor, and the biodegradation efficiency of Arthrobacter chlorophenolicus A6 on p-nitrophenol (PNP) removal in PBR at different PNP loading rates were evaluated. The novel PBR was designed to improve the hydrodynamic features such as mixing time profile (t(m95)), oxygen mass transfer coefficient (k(L)a), and overall gas hold up capacity (ɛ(G)) of the reactor. PNP concentration in the influent was varied between 600 and 1400 mg l(-1) whereas the hydraulic retention time (HRT) in the reactor was varied between 18 and 7.5h. Complete removal of PNP was achieved in the reactor up to a PNP loading rate of 2787 mg l(-1)d(-1). More than 99.9% removal of PNP was achieved in the reactor for an influent concentration of 1400 mg l(-1) and at 18 h HRT. In the present study, PNP was utilized as sole source of carbon and energy by A. chlorophenolicus A6. Furthermore, the bioreactor showed good compatibility in handling shock loading of PNP. Copyright © 2011 Elsevier B.V. All rights reserved.

Quinone-functionalized activated carbon improves the reduction of congo red coupled to the removal of p-cresol in a UASB reactor.

Science.gov (United States)

Alvarez, Luis H; Arvizu, Iris C; García-Reyes, Refugio Bernardo; Martinez, Claudia M; Olivo-Alanis, Daniel; Del Angel, Yair A

2017-09-15

In this research was immobilized anthraquinone-2-sulfonate (AQS) on granular activated carbon (GAC) to evaluate its capacity to reduce congo red (CR) in batch reactor and continuous UASB reactors. The removal of p-cresol coupled to the reduction of CR was also evaluated. Results show that the immobilization of AQS on GAC (GAC-AQS) achieved 0.469mmol/g, improving 2.85-times the electron-transferring capacity compared to unmodified GAC. In batch, incubations with GAC-AQS achieved a rate of decolorization of 2.64-fold higher than the observed with GAC. Decolorization efficiencies in UASB reactor with GAC-AQS were 83.9, 82, and 79.9% for periods I, II, and III; these values were 14.9-22.8% higher than the obtained by reactor with unmodified GAC using glucose as energy source. In the fourth period, glucose and p-cresol were simultaneously fed, increasing the decolorization efficiency to 87% for GAC-AQS and 72% for GAC. Finally, reactors efficiency decreased when p-cresol was the only energy source, but systems gradually recovered the decolorization efficiency up to 84% (GAC-AQS) and 71% (GAC) after 250 d. This study demonstrates the longest and efficient continuous UASB reactor operation for the reduction of electron-accepting contaminant in presence of quinone-functionalized GAC, but also using a recalcitrant pollutant as electron donor. Copyright © 2017 Elsevier B.V. All rights reserved.

Response of Syntrophic Propionate Degradation to pH Decrease and Microbial Community Shifts in an UASB Reactor.

Science.gov (United States)

Zhang, Liguo; Ban, Qiaoying; Li, Jianzheng; Jha, Ajay Kumar

2016-08-28

The effect of pH on propionate degradation in an upflow anaerobic sludge blanket (UASB) reactor containing propionate as a sole carbon source was studied. Under influent propionate of 2,000 mg/l and 35ºC, propionate removal at pH 7.5-6.8 was above 93.6%. Propionate conversion was significantly inhibited with stepwise pH decrease from pH 6.8 to 6.5, 6.0, 5.5, 5.0, 4.5, and then to 4.0. After long-term operation, the propionate removal at pH 6.5-4.5 maintained an efficiency of 88.5%-70.1%, whereas propionate was hardly decomposed at pH 4.0. Microbial composition analysis showed that propionate-oxidizing bacteria from the genera Pelotomaculum and Smithella likely existed in this system. They were significantly reduced at pH ≤5.5. The methanogens in this UASB reactor belonged to four genera: Methanobacterium, Methanospirillum, Methanofollis, and Methanosaeta. Most detectable hydrogenotrophic methanogens were able to grow at low pH conditions (pH 6.0-4.0), but the acetotrophic methanogens were reduced as pH decreased. These results indicated that propionate-oxidizing bacteria and acetotrophic methanogens were more sensitive to low pH (5.5-4.0) than hydrogenotrophic methanogens.

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.

2007-01-01

The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P f nl P t nl). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

Estimating the four-factor product (ε p Pfnl Ptnl) for the accurate calculation of xenon and samarium reactivities in the Syrian Miniature Neutron Source Reactor

International Nuclear Information System (INIS)

Khattab, K.

2007-01-01

The modified 135 Xe equilibrium reactivity in the Syrian Miniature Neutron Source Reactor (MNSR) was calculated first by using the WIMSD4 and CITATION codes to estimate the four-factor product (ε p P fnl P tnl ). Then, precise calculations of 135 Xe and 149 Sm concentrations and reactivities were carried out and compared during the reactor operation time and after shutdown. It was found that the 135 Xe and 149 Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149 Sm reactivities could be neglected compared to 135 Xe reactivities during the reactor operating time and after shutdown. (author)

Electrical properties of InP irradiated by fast neutrons of a nuclear reactor

International Nuclear Information System (INIS)

Kolin, N.G.; Merkurisov, D.I.; Solov'ev, S.P.

2000-01-01

Electrophysical properties of InP single crystalline samples with different initial concentration of charge carriers have been studied in relation to irradiation conditions with fast neutrons of a nuclear reactor and subsequent heat treatments within the temperature range of 20-900 deg C. It has been shown that changes of the properties depend on the initial doping level. The annealing in the temperature range mentioned above results in the elimination of radiation defects. This makes possible to use the nuclear doping method for InP samples. In this respect the contribution of intermediate neutron reactions to the whole effect of the InP nuclear doping is estimated to be ∼ 10% [ru

Evaluation of reactor induced (n,p) reactions for activation analysis of titanium in geological materials

Energy Technology Data Exchange (ETDEWEB)

Espinosa Garcia, R; Cohen, I M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1984-05-01

The possibilities of reactor induced (n,p) reactions as a tool for neutron activation analysis of titanium in geological samples are discussed. The interference of calcium and scandium is experimentally evaluated. Results for Ti, Ca and Sc in GSP-1 and PCC-1 standard rocks are presented. Based on the experimental values, it is concluded that the /sup 47/Ti(n,p)/sup 47/Sc reaction is the most favourable for titanium determination. 11 refs.

Multigroup P8 - elastic scattering matrices of main reactor elements

International Nuclear Information System (INIS)

Garg, S.B.; Shukla, V.K.

1979-01-01

To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

Review of the book: Yurkevich, G.P. Energetic reactor control systems. Editing by N.S. Khlopkin. Moscow, 2001, 344 p

International Nuclear Information System (INIS)

Kondrat'ev, V.V.

2002-01-01

Review of the book: Yurkevich, G.P. Energetic reactor control systems. Editing by N.S. Khlopkin. Moscow, 2001, is presented. Advantages of the book, specifically, easy of presentation of sophistical physical processes, new, practical and valid theoretical ways of decision of the problems are discussed. New and original decisions proposed by the author of the book are discussed. The book is beneficial for wide round of specialists [ru

Barium recovery by crystallization in a fluidized-bed reactor: effects of pH, Ba/P molar ratio and seed.

Science.gov (United States)

Su, Chia-Chi; Reano, Resmond L; Dalida, Maria Lourdes P; Lu, Ming-Chun

2014-06-01

The effects of process conditions, including upward velocity inside the column, the amount of added seed and seed size, the pH value of the precipitant or the phosphate stream and the Ba/P molar ratio in a fluidized-bed reactor (FBR) were studied with a view to producing BaHPO₄ crystals of significant size and maximize the removal of barium. XRD were used to identify the products that were collected from the FBR. Experimental results show that an upward velocity of 48 cmmin(-1) produced the largest BaHPO₄ crystals with a size of around 0.84-1.0mm. The addition of seed crystals has no effect on barium removal. The use of a seed of a size in the ranges unseededbarium was removed at pH 8.4-8.6 and [Ba]/[P]=1.0. The XRD results show that a significant amount of barium phosphate (Ba₃(PO₄)₂) was obtained at pH 11. The compounds BaHPO₄ and BaO were present at a pH of below 10. Copyright © 2014 Elsevier Ltd. All rights reserved.

Specific aspects in the manufacturing and operating of CANDU reactor pressure tubes (P/T)

International Nuclear Information System (INIS)

Muscaloiu, C.

1997-01-01

The CANDU reactor design is based on a number of individual P/T in which nuclear fuel bundles are located. P/T are required to be operated in an environment of elevated temperature (300 o C), internal pressure (10 Mpa), fast neutron flux (E>1 MeV) and heavy water. The most suitable material which can provide the desired neutron economy and still maintain its mechanical properties along with corrosion resistance is zirconium alloys Zr+ 2.5 % Nb with the following composition: niobium, 2.5 to 2.8 weight percent; oxygen, 1,000 to 1,300 ppm; zirconium + allowed impurities - balance. A total of 380 pressure tubes are installed into reactor. Each pressure tube is attached at each end to a stainless steel end fitting by means of a grooved, expanded joint. The installation works were performed by ANM Bucuresti, under the technical support of General Electric Canada. The integrity of P/T after installation was examined as follows: - the surface of the rolled area on unrolled internal surface extending 25 mm beyond rolled area was inspected for irregularities by means of a boroscope; - all pressure tubes were subjected to the helium leak test after F/C installation. During P/T operating life periodical inspections according to Canadian Standard CSA N285.4 are performed. The selection of the P/T for inspection is based either on particular properties or on the operating conditions of the fuel channel. The inspection consists in: a) Base Line Inspection within 2 years period commencing after 7,000 EFPH of operation which will include a volumetric inspection over P/T full length and measurements of P/T sag, ID, wall thickness and F/C bearing positions; b) Periodic Inspection in the same conditions plus material surveillance (on the four most significant indication P/T detected during the Base Line Inspection). The inspection will be performed on 14 selected P/T. (author)

The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

International Nuclear Information System (INIS)

Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

1977-01-01

The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

International Nuclear Information System (INIS)

Pawel, S.J.; Felde, D.K.; Pawel, R.E.

1995-10-01

To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented

Enzymatic removal of phenol and p-chlorophenol in enzyme reactor: Horseradish peroxidase immobilized on magnetic beads

International Nuclear Information System (INIS)

Bayramoglu, Guelay; Arica, M. Yakup

2008-01-01

Horseradish peroxidase was immobilized on the magnetic poly(glycidylmethacrylate-co-methylmethacrylate) (poly(GMA-MMA)), via covalent bonding and used for the treatment of phenolic wastewater in continuous systems. For this purposes, horseradish peroxidase (HRP) was covalently immobilized onto magnetic poly(GMA-MMA) beds using glutaraldehyde (GA) as a coupling agent. The maximum HRP immobilization capacity of the magnetic poly(GMA-MMA)-GA beads was 3.35 mg g -1 . The immobilized HRP retained 79% of the activity of the free HRP used for immobilization. The immobilized HRP was used for the removal of phenol and p-chlorophenol via polymerization of dissolved phenols in the presence of hydrogen peroxide (H 2 O 2 ). The effect of pH and temperature on the phenol oxidation rate was investigated. The results were compared with the free HRP, which showed that the optimum pH value for the immobilized HRP is similar to that for the free HRP. The optimum pH value for free and immobilized HRP was observed at pH 7.0. The optimum temperature for phenols oxidation with immobilized HRP was between 25 and 35 deg. C and the immobilized HRP has more resistance to temperature inactivation than that of the free form. Finally, the immobilized HRP was operated in a magnetically stabilized fluidized bed reactor, and phenols were successfully removed in the enzyme reactor

Towards nuclear fusion reactors

International Nuclear Information System (INIS)

1993-11-01

The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

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

International Nuclear Information System (INIS)

Cheng Shouyu; Peng Minjun; Liu Xinkai

2014-01-01

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

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Exploring the selective lactic acid production from food waste in uncontrolled pH mixed culture fermentations using different reactor configurations.

Science.gov (United States)

Bonk, Fabian; Bastidas-Oyanedel, Juan-Rodrigo; Yousef, Ahmed F; Schmidt, Jens Ejbye

2017-08-01

Carboxylic acid production from food waste by mixed culture fermentation is an important future waste management option. Obstacles for its implementation are the need of pH control, and a broad fermentation product spectrum leading to increased product separation costs. To overcome these obstacles, the selective production of lactic acid (LA) from model food waste by uncontrolled pH fermentation was tested using different reactor configurations. Batch experiments, semi-continuously fed reactors and a percolation system reached LA concentrations of 32, 16 and 15gCOD LA /L, respectively, with selectivities of 93%, 84% and 75% on COD base, respectively. The semi-continuous reactor was dominated by Lactobacillales. Our techno-economic analysis suggests that LA production from food waste can be economically feasible, with LA recovery and low yields remaining as major obstacles. To solve both problems, we successfully applied in-situ product extraction using activated carbon. Copyright © 2017 Elsevier Ltd. All rights reserved.

Novel p38α MAP kinase inhibitors identified from yoctoReactor DNA-encoded small molecule library

DEFF Research Database (Denmark)

Petersen, L. K.; Blakskjær, P.; Chaikuad, A.

2016-01-01

A highly specific and potent (7 nM cellular IC50) inhibitor of p38α kinase was identified directly from a 12.6 million membered DNA-encoded small molecule library. This was achieved using the high fidelity yoctoReactor technology (yR) for preparing the DNA-encoded library, and a homogeneous...... interactions. Moreover, the crystal structure showed, that although buried in the p38α active site, the original DNA attachment point of the compound was accessible through a channel created by the distorted P-loop conformation. This study demonstrates the usability of DNA-encoded library technologies...

Syngas fermentation by Clostridium carboxidivorans P7 in a horizontal rotating packed bed biofilm reactor with enhanced ethanol production

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert C.; Wen, Zhiyou

2017-01-01

Highlights: • A novel a horizontal rotating packed bed (h-RPB) reactor for syngas fermentation was reported. • The h-RPB reactor enhanced ethanol productivity by 3.3-folds compared to continuous stirred tank reactor (CSTR). • The h-RPB reactor has a unique feature of transfer gas from both bulk liquid phase and headspace phase. • The mass transfer in the headspace of h-PRB played an important role for enhanced ethanol production. - Abstract: Gasification of lignocellulosic biomass followed by syngas fermentation is a promising process for producing fuels and chemicals. Syngas fermentation, however, is commonly limited by low mass transfer rates. In this work, a horizontally oriented rotating packed bed (h-RPB) reactor was developed to improve mass transfer and enhance ethanol production. In the h-RPB reactor, cell attachment materials were packed in the reactor and half submerged in the liquid and half exposed to the headspace. With continuous rotation of the packing materials, the cells in biofilm were alternately in contact with liquid and headspace; thus, transport of syngas to the cells occurred in both the liquid phase and headspace. The volumetric mass transfer coefficient (k_La) of the h-RPB reactor was lower than that in a traditional continuous stirred tank reactor (CSTR), indicating the mass transfer in the liquid phase of h-PRB was lower than CSTR, and the mass transfer in the headspace phase played an important role in syngas fermentation. The syngas fermentation of Clostridium carboxidivorans P7 in h-RPB resulted in a 7.0 g/L titer and 6.7 g/L/day productivity of ethanol, respectively, 3.3 times higher than those obtained in a CSTR under the same operational conditions. The results demonstrate that the h-RPB reactor is an efficient system for syngas fermentation, making cellulosic ethanol biorefinery one step closer to technical and economic feasibility.

Reactor production and separation of no-carrier added 32P for medical applications

International Nuclear Information System (INIS)

Vimalnath, K.V.; Shetty, P.; Rajeswari, A.; Chirayil, V.; Chakraborty, S.; Dash, A.

2014-01-01

diluted samples of 32 P of each batch were analyzed by gamma ray spectroscopy. Radionuclide purity as determined by paper chromatography indicates >98% of 32 P as orthophosphate ion. The levels of trace metal ions as analyzed by ICP-AES were found to be below detectable limit (0.1 mg/L, 0.1 ppm for all the metal ions) thereby confirming their absence. Using the described procedure, several batches of 32 P were prepared and results of 32 P production carried out from 5 typical batches are presented. The data represents the typical yields of the 32 P in regular batches. The relatively larger variations of yield from batch to batch are mostly due to the differences in the reactor irradiation conditions, such as the exact duration, intervening shut-down, and the variation of fast neutron flux level due to the power level of the reactor operation. The objective of separating 32 P from neutron irradiated sulphur for in vivo application has been successfully accomplished

Magnetite synthesis from ferrous iron solution at pH 6.8 in a continuous stirred tank reactor.

Science.gov (United States)

Mos, Yvonne M; Zorzano, Karin Bertens; Buisman, Cees J N; Weijma, Jan

2018-04-01

Partial oxidation of defined Fe 2+ solutions is a well-known method for magnetite synthesis in batch systems. The partial oxidation method could serve as basis for an iron removal process in drinking water production, yielding magnetite (Fe 3 O 4 ) as a compact and valuable product. As a first step toward such a process, a series of experiments was carried out, in which magnetite was synthesized from an Fe 2+ solution in a 2 L continuous stirred tank reactor (CSTR) at atmospheric pressure and 32 °C. In four experiments, elevating the pH from an initial value of 5.5 or 6.0 to a final value of 6.8, 7.0 or 7.5 caused green rust to form, eventually leading to magnetite. Formation of NH 4 + in the reactor indicated that NO 3 - and subsequently NO 2 - served as the oxidant. However, mass flow analysis revealed an influx of O 2 to the reactor. In a subsequent experiment, magnetite formation was achieved in the absence of added nitrate. In another experiment, seeding with magnetite particles led to additional magnetite precipitation without the need for a pH elevation step. Our results show, for the first time, that continuous magnetite formation from an Fe 2+ solution is possible under mild conditions, without the need for extensive addition of chemicals.

Analysis of trichloroethylene removal and bacterial community function based on pH-adjusted in an upflow anaerobic sludge blanket reactor.

Science.gov (United States)

Zhang, Ying; Hu, Miao; Li, Pengfei; Wang, Xin; Meng, Qingjuan

2015-11-01

The study reported the upflow anaerobic sludge blanket (UASB) reactor performance in treating wastewater containing trichloroethylene (TCE) and characterized variations of bacteria composition and structure by changing the pH from 6.0 to 8.0. A slightly acidic environment (pH < 7.0) had a greater impact on the TCE removal. Illumina pyrosequencing was applied to investigate the bacterial community changes in response to pH shifts. The results demonstrated that pH greatly influenced the dominance and presence of specific populations. The potential TCE degradation pathway in the UASB reactor was proposed. Importantly, the genus Dehalobacter which was capable of reductively dechlorinating TCE was detected, and it was not found at pH of 6.0, which presumably is the reason why the removal efficiency of TCE was the lowest (80.73 %). Through Pearson correlation analyses, the relative abundance of Dehalobacter positively correlated with TCE removal efficiency (R = 0.912). However, the relative abundance of Lactococcus negatively correlated with TCE removal efficiency according to the results from Pearson correlation analyses and redundancy analysis (RDA).

Feasibility studies on the production of essential radioisotopes (24Na and 32P) using the Ghana Miniature Neutron Source Reactor (GHARR-1)

International Nuclear Information System (INIS)

Dotse, S.C.

2012-01-01

Feasibility studies on the production of 32 P and 24 Na using a Miniature Neutron Source Reactor named Ghana Research Reactor-1 (GHARR-1) has been conducted. A theoretical model of the cyclic activation technique was developed for the simulation of specific activities under varying parameters. Specific activity values estimated for four cycles of irradiation with activation parameters falling within the specifications of the reactor were experimentally validated. Experimented results were compared to those theoretically estimated for both 24 Na and 32 P. Experimented specific activity values for both radioisotopes generally fell below their theoretical values but recorded activity build-ups from one cycle to the other. The 24 Na nuclide showed a regular pattern for the increase in specific activity from one cycle to the next with an average percentage difference of experimented to theoretical value of 19%. The 32 P nuclide showed an irregular pattern for the increase in specific activity from one cycle to the next with an average percentage difference of experimental to theoretical value of 11%. The specific activities experimentally attained, with reference to activity levels used for various applications in agriculture and industry suggests the cyclic activation technique can be used for the production of radioisotopes of appreciable activities using low power research reactors, which are characterised with limited excess core reactivity and cannot sustain long periods of irradiation. (au)

Compliance of the Savannah River Plant P-Reactor cooling system with environmental regulations. Demonstrations in accordance with Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972

International Nuclear Information System (INIS)

Wilde, E.W.

1985-12-01

This document presents demonstrations under Sections 316(a) and (b) of the Federal Water Pollution Control Act of 1972 for the P-Reactor cooling system at the Savannah River Plant (SRP). The demonstrations were mandated when the National Pollution Discharge Elimination System (NPDES) permit for the SRP was renewed and the compliance point for meeting South Carolina Class B water quality criteria in the P-Reactor cooling system was moved from below Par Pond to the reactor cooling water outfall, No. P-109. Extensive operating, environmental, and biological data, covering most of the current P-Reactor cooling system history from 1958 to the present are discussed. No significant adverse effects were attributed to the thermal effluent discharged to Par Pond or the pumping of cooling water from Par Pond to P Reactor. It was conluded that Par Pond, the principal reservoir in the cooling system for P Reactor, contains balanced indigenous biological communities that meet all criteria commonly used in defining such communities. Par Pond compares favorably with all types of reservoirs in South Carolina and with cooling lakes and reservoirs throughout the southeast in terms of balanced communities of phytoplankton, macrophytes, zooplankton, macroinvertebrates, fish, and other vertebrate wildlife. The report provides the basis for negotiations between the South Carolina Department of Health and Environmental Control (SCDHEC) and the Department of Energy - Savannah River (DOE-SR) to identify a mixing zone which would relocate the present compliance point for Class B water quality criteria for the P-Reactor cooling system

Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

Energy Technology Data Exchange (ETDEWEB)

Chen, Wei-Yu; Wu, Jer-Horng, E-mail: enewujh@mail.ncku.edu.tw; Lin, Shih-Chiang; Chang, Juu-En

2016-07-15

Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

Bioremediation of polychlorinated-p-dioxins/dibenzofurans contaminated soil using simulated compost-amended landfill reactors under hypoxic conditions

International Nuclear Information System (INIS)

Chen, Wei-Yu; Wu, Jer-Horng; Lin, Shih-Chiang; Chang, Juu-En

2016-01-01

Highlights: • We developed a new hypoxic reactor system for remediating PCDD/Fs. • We demonstrated effects of compost on the degradation of PCDD/Fs. • We uncovered microbial compositions and dynamics during the degradation of PCDD/Fs. - Abstract: Compost-amended landfill reactors were developed to reduce polychlorinated-p-dioxins and dibenzofurans (PCDD/Fs) in contaminated soils. By periodically recirculating leachate and suppling oxygen, the online monitoring of the oxidation reduction potential confirmed that the reactors were maintained under hypoxic conditions, with redox levels constantly fluctuating between −400 and +80 mV. The subsequent reactor operation demonstrated that PCDD/F degradation in soil could be facilitated by amending compost originating from the cow manure and waste sludge and that the degradation might be affected by the availability of easily degradable substrates in the soil and compost. The pyrosequencing analysis of V4/V5 regions of bacterial 16S rRNA genes suggested that species richness of the soil microbial community was increased by a factor of 1.37–1.61. Although the bacterial community varied with the compost origin and changed markedly during reactor operation, it was dominated by Alphaproteobacteria, Gammaproteobacteria, Actinobacteria, and Firmicutes. The aerotolerant anaerobic Sedimentibacter and Propionibacterium spp., and the uncultured Chloroflexi group could be temporarily induced to a high abundance by amending the cow manure compost; the bacterial growths were associated with the rapid degradation of PCDD/Fs. Overall, the novel bioremediation method for PCDD/F-contaminated soils using hypoxic conditions was effective, simple, energy saving, and thus easily practicable.

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

type of reactors: low power reactors, graphite and natural uranium reactors, heavy water and natural uranium reactors (with water, gas or heavy water as coolant), heavy water and enriched uranium reactors, 'kettle' type reactors, pool type reactors and high power reactors. Finally, additional factors are considered as the correlation of the reactor type with the research program, safety considerations, size of the construction and economical considerations. (M.P.)

MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

Energy Technology Data Exchange (ETDEWEB)

Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1968-06-15

Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

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

DEFF Research Database (Denmark)

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

2012-01-01

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

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.)

SIRIUS-P: An inertially confined direct drive laser fusion power reactor

International Nuclear Information System (INIS)

Sviatoslavsky, I.N.; Kulcinski, G.L.; Moses, G.A.; Bruggink, D.; Engelstad, R.L.; Khater, H.Y.; Larsen, E.M.; Lovell, E.G.; MacFarlane, J.J.; Mogahed, E.A.; Peterson, R.R.; Sawan, M.E.; Wang, P.; Wittenberg, L.J.

1993-03-01

The SIRIUS-P conceptual design study is of a 1000 MWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. The reference driver is a KrF laser; however, any other laser capable of delivering short wavelength energy can be substituted. Sixty beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz and a target gain of 118. The spherical chamber has an internal diameter of 6.5 m and consists of two independent components, a first wall assembly fabricated from a c/c composite and a blanket assembly made of SiC. First wall protection is provided by a xenon buffer gas at a pressure of 0.5 torr. The chamber is cooled by a flowing granular bed of solid ceramic material, TiO 2 for the first wall assembly and Li 2 O for the blanket assembly. The chamber is housed within a 42 m radius cylindrical reactor building which is 86 m high and which shares the same vacuum space as the chamber. All the laser beams are brought in at the bottom of the building, first onto a dielectrically coated final focusing mirror and finally onto a metallic grazing incidence mirror which reflects them into the chamber through beam ports open to the building. Neutron traps behind the grazing incidence mirrors are used to prolong the lifetimes of the final focusing optics. The nominal cost of electricity from this system is 65 mills/kwh assuming an 8% interest rate on capital

Nuclear reactor theory

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2007-09-01

This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Development task of compact reactor

International Nuclear Information System (INIS)

Kurushima, Morihiro

1982-01-01

In the Ministry of International Trade and Industry, studies proceed on the usage of compact medium and small LWRs. As such, the reactors from 100 to 200 MW may meet varieties of demands in scale and kind in view of the saving of petroleum and the economy of nuclear power. In this case, the technology of light water reactors with already established safety will be suitable for the development of compact reactors. The concept of ''nuclear power community'' using the compact reactors in local society and industrial zones was investigated. The following matters are described: need for the introduction of compact reactors, the survey on the compact reactor systems, and the present status and future problems for compact reactor usage. (J.P.N.)

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

DEFF Research Database (Denmark)

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

2012-01-01

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

Neutronic study of the two french heavy water reactors

International Nuclear Information System (INIS)

Horowitz, J.

1955-01-01

The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

Establishment of regulatory framework for the development reactor licensing

International Nuclear Information System (INIS)

Jo, Jong C.; Yune, Young G.; Kim, Woong S.; Ahn, Sang K.; Kim, In G.; Kim, Hho J.

2003-01-01

With a trend that various types of advanced reactor designs are currently under development worldwide, the Korea Atomic Energy Research Institute has been developing an advanced reactor called ' System-integrated Modular Advanced Reactor (SMART)', which is a small sized integral type pressurized water reactor with a rated thermal power of 330 MW. To demonstrate the safety and the performance of the SMART reactor design, the SMART Research and Development Center has embarked to build a scaled-down pilot plant of SMART, called 'SMART-P' with a rated thermal power of 65 MW. In preparation for the forthcoming applications for both construction permit and operating license of SMART-P in the near future, the Korea Institute of Nuclear Safety is developing a new regulatory framework for licensing review of such a development reactor, which covers establishment of licensing process, identification and resolution of technical and safety issues, development of regulatory evaluation or verification-purpose computer codes and analytical methods, and establishment of design-specific, general design and operating criteria, regulations, and associated regulatory guides. This paper presents the current activities for establishing a regulatory framework for the licensing of a research and development reactor. Discussions are made on the SMART-P development program, the current Korean regulatory framework for reactor licensing, the SMART-P licensing-related issues, and the approach and strategy for developing an effective regulatory framework for the SMART-P licensing

Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

International Nuclear Information System (INIS)

Carey, W.M.; Anderson, T.T.; Bobis, J.P.

1976-06-01

Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

Structural analysis of the P reactor at the Savannah River Site

International Nuclear Information System (INIS)

Zaslawsky, M.; Maryak, M.

1991-01-01

A seismic analysis of the P-reactor buildings that were built in the early 1950's has been performed using current criteria and analysis techniques. The seismic input is based on the RG 1.60 free field response spectra anchored at 0.2g ZPA. The SSI analysis applied deconvolution techniques to establish soil parameters based on strain dependent damping and shear modulus relationships. The analysis used 2-dimensional soil structure interaction techniques to generate floor response spectra. The spectra were adjusted to account for torsional amplifications resulting from differences between the locations of the center of mass of the floors and the center of rigidity of the connecting vertical column elements. The resulting floor response spectra were smoothed and broadened in accordance with NRC criteria. In addition to developing floor response spectra, building shears and moments were obtained and an assessment of the structural capacity of the buildings to withstand the seismic loads was made

Environmental consequences of alternatives to L Reactor restart

International Nuclear Information System (INIS)

1983-01-01

Alternatives to renewed L-Reactor operation for increased production of nuclear materials are: restart of R Reactor, construction and operation of a New Production Reactor (NPR), increased throughput of SRP reactors C, K, and P and N Reactor at Hanford, restart of K Reactors at Hanford, and no action - standby ready state for L Reactor. This report compares the environmental consequences from the proposed L-Reactor restart and these alternatives. The environmental consequences considered are radiological releases, radiocesium remobilization, nonradiological releases, ecological impacts and transportation

Phosphorus removal from UASB reactor effluent by reactive media filtration.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno

2017-08-01

The phosphorus (P) and BOD7 removal performance of an upflow packed bed reactor (PBR) filled with two reactive filter media was studied over 50 weeks. The lower one-fifth of the reactor was filled with calcium-silicate-hydrate (Sorbulite®) and the upper four-fifth with calcium-silicate (Polonite®). A laboratory-scale upflow anaerobic sludge bed reactor (UASB) delivered wastewater to the PBR. A model was developed to describe the gradient in P concentration change in the reactor, based on reaction kinetics. The reaction terms were assumed to follow the Langmuir isotherm, based on the results obtained in a batch test. First, a comparison was made between experimental and simulated results. The capability of the model to forecast P removal capacity was then tested for three hypothetical cases: (i) reactor filled with Sorbulite and Polonite, (ii) reactor filled with only Sorbulite, and (iii) reactor filled with only Polonite. Finally, a sensitivity analysis was performed for the main parameters in the model. The average removal of P and BOD7 from the UASB effluent was 98% and 90%, respectively. The starting pH of the dual-medium effluent was 12.2 and decreased gradually over time to 11.1. The simulation both overestimated and underestimated mean measured P removal but was within the range of maximum and minimum measured values. The hypothetical cases revealed that most P was removed by Polonite due to calcium phosphate precipitation. The removal capacity of the two filter materials and their layer height in the reactor were the most sensitive parameters in the simulation.

Enhancing mass transfer and ethanol production in syngas fermentation of Clostridium carboxidivorans P7 through a monolithic biofilm reactor

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert; Wen, Zhiyou

2014-01-01

Highlights: • Syngas fermentation process is limited by gas-to-liquid mass transfer. • A novel monolithic biofilm reactor (MBR) for efficient mass transfer was developed. • MBR with slug flow resulted in higher k L a than bubble column reactor (BCR). • MBR enhanced ethanol productivity by 53% compared to BCR. • MBR was demonstrated as a promising reactor configuration for syngas fermentation. - Abstract: Syngas fermentation is a promising process for producing fuels and chemicals from lignocellulosic biomass. Currently syngas fermentation faces several engineering challenges, with gas-to-liquid mass transfer limitation representing the major bottleneck. The aim of this work is to evaluate the performance of a monolithic biofilm reactor (MBR) as a novel reactor configuration for syngas fermentation. The volumetric mass transfer coefficient (k L a) of the MBR was evaluated in abiotic conditions within a wide range of gas flow rates (i.e., gas velocity in monolithic channels) and liquid flow rates (i.e., liquid velocity in the channels). The k L a values of the MBR were higher than those of a controlled bubble column reactor (BCR) in certain conditions, due to the slug flow pattern in the monolithic channels. A continuous syngas fermentation using Clostridium carboxidivorans P7 was conducted in the MBR system under varying operational conditions, with the variables including syngas flow rate, liquid recirculation between the monolithic column and reservoir, and dilution rate. It was found that the syngas fermentation performance – measured by such parameters as syngas utilization efficiency, ethanol concentration and productivity, and ratio of ethanol to acetic acid – depended not only on the mass transfer efficiency but also on the biofouling or abrading of the biofilm attached on the monolithic channel wall. At a condition of 300 mL/min of syngas flow rate, 500 mL/min of liquid flow rate, and 0.48 day −1 of dilution rate, the MBR produced much higher

Petascale algorithms for reactor hydrodynamics

International Nuclear Information System (INIS)

Fischer, P.; Lottes, J.; Pointer, W.D.; Siegel, A.

2008-01-01

We describe recent algorithmic developments that have enabled large eddy simulations of reactor flows on up to P = 65, 000 processors on the IBM BG/P at the Argonne Leadership Computing Facility. Petascale computing is expected to play a pivotal role in the design and analysis of next-generation nuclear reactors. Argonne's SHARP project is focused on advanced reactor simulation, with a current emphasis on modeling coupled neutronics and thermal-hydraulics (TH). The TH modeling comprises a hierarchy of computational fluid dynamics approaches ranging from detailed turbulence computations, using DNS (direct numerical simulation) and LES (large eddy simulation), to full core analysis based on RANS (Reynolds-averaged Navier-Stokes) and subchannel models. Our initial study is focused on LES of sodium-cooled fast reactor cores. The aim is to leverage petascale platforms at DOE's Leadership Computing Facilities (LCFs) to provide detailed information about heat transfer within the core and to provide baseline data for less expensive RANS and subchannel models.

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 %).

Alkaline cyanide degradation by Pseudomonas pseudoalcaligenes CECT5344 in a batch reactor. Influence of pH

International Nuclear Information System (INIS)

Huertas, M.J.; Saez, L.P.; Roldan, M.D.; Luque-Almagro, V.M.; Martinez-Luque, M.; Blasco, R.; Castillo, F.; Moreno-Vivian, C.; Garcia-Garcia, I.

2010-01-01

Water containing cyanide was biologically detoxified with the bacterial strain Pseudomonas pseudoalcaligenes CECT5344 in a batch reactor. Volatilization of toxic hydrogen cyanide (HCN) was avoided by using an alkaline medium for the treatment. The operational procedure was optimized to assess cyanide biodegradation at variable pH values and dissolved oxygen concentrations. Using an initial pH of 10 without subsequent adjustment allowed total cyanide to be consumed at a mean rate of approximately 2.81 mg CN - L -1 O.D. -1 h -1 ; however, these conditions posed a high risk of HCN formation. Cyanide consumption was found to be pH-dependent. Thus, no bacterial growth was observed with a controlled pH of 10; on the other hand, pH 9.5 allowed up to 2.31 mg CN - L -1 O.D. -1 h -1 to be converted. The combination of a high pH and a low dissolved oxygen saturation (10%) minimized the release of HCN. This study contributes new basic knowledge about this biological treatment, which constitutes an effective alternative to available physico-chemical methods for the purification of wastewater containing cyanide or cyano-metal complexes.

Physical aspects of liquid-impelled loop reactors

NARCIS (Netherlands)

Sonsbeek, van H.

1992-01-01

<p>The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

Reactor power control system

International Nuclear Information System (INIS)

Tomisawa, Teruaki.

1981-01-01

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

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

International Nuclear Information System (INIS)

Essadki, A.H.; Gourich, B.; Vial, Ch.; Delmas, H.; Bennajah, M.

2009-01-01

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm 2 , but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H 2 microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-08-15

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Reactor incident status 1981 annual report

International Nuclear Information System (INIS)

Kiser, S.H.

1982-01-01

Reactor Incident followup action is summarized through periodic status reports. This annual report summarizes action taken or anticipated for Reactor Incidents through December 1981. Incidents for which action has been completed, have been deleted from the report. Quarterly addende will update the report by tabulating incidents for each three month period through the coming year. The report consists of a part for the P, K, and C Reactors. Each reactor part is divided into three sections: Further Technical Analysis or Followup Needed; Funding and/or Implementation Needed; and No Further Technical Analysis Anticipated

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)

Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

International Nuclear Information System (INIS)

Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

1995-01-01

Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system

P R Vasudeva Rao

Indian Academy of Sciences (India)

Home; Journals; Sadhana. P R Vasudeva Rao. Articles written in Sadhana. Volume 27 Issue 5 October 2002 pp 527-558. Development of fuels and structural materials for fast breeder reactors · Baldev Raj S L Mannan P R Vasudeva Rao M D Mathew · More Details Abstract Fulltext PDF. Fast breeder reactors (FBRs) are ...

Economic targets for small PWR reactor designs

International Nuclear Information System (INIS)

Board, J.

1991-01-01

Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

The secure heating reactor

International Nuclear Information System (INIS)

Pind, C.

1987-01-01

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives for heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories

Low cost UV-Ozone reactor mounted for treatment of electrode anodes used in P-OLEDs devices

Directory of Open Access Journals (Sweden)

Emerson Roberto Santos

Full Text Available Abstract Low cost UV-Ozone reactor using a high pressure mercury vapor lamp of 80 watts without outer bulb showed good results for treatment of ITO films used as anode electrode in the assembly of P-OLED (polymer-organic light emitting diode devices. This study revealed 20 minutes as effective treatment time and it was verified also that the effect of UV-Ozone treatment loses its efficiency as the elapsed time increases. It was analyzed with measurements of contact angle using a droplet of PEDOT:PSS polymer. P-OLEDs devices were mounted with architecture: ITO/PEDOT:PSS/PVK/Alq3/Al. The PVK polymer was diluted in organic solvent of 1,2,4-trichlorobenzene with concentrations of: 5, 10, 20 and 30 mg/mL. Results revealed better performance of P-OLED devices for concentration of 5 mg/mL resulting in lower threshold voltage, elevation of electrical current and similar diode curve.

Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

Energy Technology Data Exchange (ETDEWEB)

Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

2009-09-15

Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

Alkaline cyanide degradation by Pseudomonas pseudoalcaligenes CECT5344 in a batch reactor. Influence of pH

Energy Technology Data Exchange (ETDEWEB)

Huertas, M.J., E-mail: mjhuertas@us.es [Instituto de Bioquimica Vegetal y Fotosintesis, CSIC-Universidad de Sevilla Avda Americo Vespucio, 49, 41092 Sevilla (Spain); Saez, L.P.; Roldan, M.D.; Luque-Almagro, V.M.; Martinez-Luque, M. [Departamento de Bioquimica y Biologia Molecular, Edificio Severo Ochoa, 1a Planta, Campus de Rabanales, Universidad de Cordoba, 14071 Cordoba (Spain); Blasco, R. [Departamento de Bioquimica y Biologia Molecular y Genetica, Facultad de Veterinaria, Universidad de Extremadura, 11071 Caceres (Spain); Castillo, F.; Moreno-Vivian, C. [Departamento de Bioquimica y Biologia Molecular, Edificio Severo Ochoa, 1a Planta, Campus de Rabanales, Universidad de Cordoba, 14071 Cordoba (Spain); Garcia-Garcia, I. [Departamento de Ingenieria Quimica, Edificio Marie Curie, Campus de Rabanales, Universidad de Cordoba, 14071 Cordoba (Spain)

2010-07-15

Water containing cyanide was biologically detoxified with the bacterial strain Pseudomonas pseudoalcaligenes CECT5344 in a batch reactor. Volatilization of toxic hydrogen cyanide (HCN) was avoided by using an alkaline medium for the treatment. The operational procedure was optimized to assess cyanide biodegradation at variable pH values and dissolved oxygen concentrations. Using an initial pH of 10 without subsequent adjustment allowed total cyanide to be consumed at a mean rate of approximately 2.81 mg CN{sup -} L{sup -1} O.D.{sup -1} h{sup -1}; however, these conditions posed a high risk of HCN formation. Cyanide consumption was found to be pH-dependent. Thus, no bacterial growth was observed with a controlled pH of 10; on the other hand, pH 9.5 allowed up to 2.31 mg CN{sup -} L{sup -1} O.D.{sup -1} h{sup -1} to be converted. The combination of a high pH and a low dissolved oxygen saturation (10%) minimized the release of HCN. This study contributes new basic knowledge about this biological treatment, which constitutes an effective alternative to available physico-chemical methods for the purification of wastewater containing cyanide or cyano-metal complexes.

Advanced designs of VVER reactor plant

International Nuclear Information System (INIS)

Mokhov, V.A.

2010-01-01

The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

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

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

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

Oxidation of aluminum alloy cladding for research and test reactor fuel

Science.gov (United States)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Oxidation of aluminum alloy cladding for research and test reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2008-08-31

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Localized corrosion problems in water reactors

International Nuclear Information System (INIS)

Coriou, Henri.

1977-01-01

Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

[Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

Science.gov (United States)

Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

2010-03-01

Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

Implantation of β-emitters on biomedical implants: 32 P isotropic ion implantation using a coaxial plasma reactor

International Nuclear Information System (INIS)

Fortin, M.A.; Paynter, R.W.; Sarkissian, A.; Stansfield, B.L.; Terreault, B.; Dufresne, V.

2003-01-01

The development of endovascular brachytherapy and the treatment of certain types of cancers (liver, lung, prostate) often require the use of beta-emitters, sometimes in the form of radioisotope-implanted devices. Among the most commonly used isotopes figures 32 P, a pure beta-emitter (maximum energy: 1.7 MeV), of which the path in biological tissues is of a few cm, restricting the impact of electron bombardment to the immediate environment of the implant. Several techniques and processes have been tried to elaborate surfaces and devices showing strongly bonded, or implanted 32 P. Anodizing, vapor phase deposition, grafting of oligonucleotides, as well as ion implantation processes have been investigated by several research groups as methods to implant beta-radioisotopes into surfaces. A coaxial plasma reactor was developed at INRS to implant radioisotopes into cylindrical metallic objects, such as coronary stents commonly used in angioplasty procedures. The dispersion of 32 P atoms on the interior surfaces of the chamber can be investigated using radiographs, contributing to image the plasma ion transport mechanisms that guide the efficiency of the implantation procedure. The amount of radioactivity on the wall liner, on the internal components, and on the biomedical implants are quantified using a surface barrier detector. A comparative study establishes a relationship between the gray scale of the radiographs, and dose measurements. A program was developed to convert the digitized images into maps showing surface dose density in mCi/cm 2 . An integration process allows the quantification of the doses on the walls and components of the reactor. Finally, the resulting integral of the 32 P dose is correlated to the initial amount of radioactivity inserted inside the implanter before the dismantling procedure. This method could be introduced as a fast and reliable way to test, qualify and assess the amount of radioactivity present on the as-produced implants

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

Energy Technology Data Exchange (ETDEWEB)

Sponza, Delia Teresa, E-mail: delya.sponza@deu.edu.tr [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: pinar.demirden@kozagold.com [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)

2010-04-15

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

International Nuclear Information System (INIS)

Sponza, Delia Teresa; Demirden, Pinar

2010-01-01

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

Diagnostic measurement on research reactors

International Nuclear Information System (INIS)

Dach, K.; Zbytovsky, A.

A comparison is made of noise experiments on zero power and power reactors. The general characteristics of noise experiments on power reactors is their ''passivity'', i.e., the experiment does not require any interruption of the normal operating regime of the reactor system. On zero power research reactors where the fission reaction constitutes the dominant noise source such conditions have to be created in the study of noise components as to make the investigated noise dominant and the noise of the fission reaction the background. The simultaneous use of both methods makes it possible to determine the spectral composition of reactivity fluctuations, which facilitates the identification of noise sources. The conditions are described of the recordability of noise components. The possibilities are listed provided for research work in Czechoslovakia and the possibility is studied of setting up an expert team to organize the respective experimental programme on an international scale. Power reactors manufactured in the GDR are considered as the suitable experimental base. (J.P.)

Complete automation of nuclear reactors control

International Nuclear Information System (INIS)

Weill, J.

1955-01-01

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

Seismic fragility analysis of buried steel piping at P, L, and K reactors

International Nuclear Information System (INIS)

Wingo, H.E.

1989-10-01

Analysis of seismic strength of buried cooling water piping in reactor areas is necessary to evaluate the risk of reactor operation because seismic events could damage these buried pipes and cause loss of coolant accidents. This report documents analysis of the ability of this piping to withstand the combined effects of the propagation of seismic waves, the possibility that the piping may not behave in a completely ductile fashion, and the distortions caused by relative displacements of structures connected to the piping

Influence of P-Reactor operation on the aquatic ecology of Par Pond: a literature review

International Nuclear Information System (INIS)

Wilde, E.W.; Tilly, L.J.

1985-02-01

Par Pond is a 1012 hectare reservoir that was constructed in 1958 to provide cooling water for Savannah River nuclear reactors. The purpose of this report is to summarize all known studies on the Par Pond system and point out demonstrable or probable effects that can be correlated with reactor operations. Reactor operation effects the Par Pond ecosystem through: (1) pumping, (2) thermal alteration, and (3) the addition of Savannah River makeup water. The influence of each of these factors is discussed. 108 references, 24 figures, 34 tables. (MF)

Current safety issues related to research reactor operation

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

2000-01-01

The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (author)

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 shutdown: nuclear power plant performance

International Nuclear Information System (INIS)

Anon.

1982-01-01

The article essentially looks at the performance of nine of Sweden's nuclear reactors. A table lists the percentage of time for the first three quarters of 1981 that the reactors were operating, and the number of hours out of service for planned or other reasons. In particular, one station - Ringhals 3 - was out of action because of a damaged tube in the associated steam generator. The same fault occurred with another reactor - Ringhals 4 - before this was brought into service. The reasons for the failure and its importance are briefly discussed. (G.P.)

Contour analysis of steady state tokamak reactor performance

International Nuclear Information System (INIS)

Devoto, R.S.; Fenstermacher, M.E.

1990-01-01

A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

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.)

A multi-agent design for a pressurized water reactor (P.W.R.) control system

International Nuclear Information System (INIS)

Aimar-Lichtenberger, M.

1999-01-01

This PhD work is in keeping with the complex industrial process control. The starting point is the analysis of control principles in a Pressurized Water Reactor (P.W.R). In order to cope with the limits of the present control procedures, a new control organisation by objectives and means is defined. This functional organisation is based on the state approach and is characterized by the parallel management of control functions to ensure the continuous control of the installation essential variables. With regard to this complex system problematic, we search the most adapted computer modeling. We show that a multi-agent system approach brings an interesting answer to manage the distribution and parallelism of control decisions and tasks. We present a synthetic study of multi-agent systems and their application fields.The choice of a multi-agent approach proceeds with the design of an agent model. This model gains experiences from other applications. This model is implemented in a computer environment which combines the mechanisms of an object language with Prolog. We propose in this frame a multi-agent modeling of the control system where each function is represented by an agent. The agents are structured in a hierarchical organisation and deal with different abstraction levers of the problem. Following a prototype process, the validation is realized by an implementation and by a coupling to a reactor simulator. The essential contributions of an agent approach turn on the mastery of the system complexity, the openness, the robustness and the potentialities of human-machine cooperation. (author)

Development of technologies for nuclear reactors of small and medium sized

International Nuclear Information System (INIS)

2011-08-01

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Production of energy in a high temperature reactor

International Nuclear Information System (INIS)

Anon.

1979-01-01

The cooling gas having left the reactor core is fed to a generator for direct production of current from the kinetic energy. Afterwards the gas is fed to a heat exchanger for cooling, then compressed and refed to the reactor core. The method further comprises that one part of the energy of the fission material is directly converted to electric energy in the reactor core, whereas the other part of the energy of the fission material is impressed upon the cooling gas. According to the invention the cooling gas when entering the reactor is first fed to that part of the reactor core which serves as a thermoionic or thermoelectric transducer. Afterwards the cooling gas is fed to the remaining part of the reactor gas. (P.K.)

Thermal energy and bootstrap current in fusion reactor plasmas

International Nuclear Information System (INIS)

Becker, G.

1993-01-01

For DT fusion reactors with prescribed alpha particle heating power P α , plasma volume V and burn temperature i > ∼ 10 keV specific relations for the thermal energy content, bootstrap current, central plasma pressure and other quantities are derived. It is shown that imposing P α and V makes these relations independent of the magnitudes of the density and temperature, i.e. they only depend on P α , V and shape factors or profile parameters. For model density and temperature profiles analytic expressions for these shape factors and for the factor C bs in the bootstrap current formula I bs ∼ C bs (a/R) 1/2 β p I p are given. In the design of next-step devices and fusion reactors, the fusion power is a fixed quantity. Prescription of the alpha particle heating power and plasma volume results in specific relations which can be helpful for interpreting computer simulations and for the design of fusion reactors. (author) 5 refs

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.)

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.)

Eh-pH diagrams for elements from Z = 40 to Z = 52: application to the Oklo natural reactor, Gabon

Energy Technology Data Exchange (ETDEWEB)

Brookins, D G [New Mexico Univ., Albuquerque (USA). Dept. of Geology

1978-11-01

Eh-pH diagrams for elements from Z = 40 (Zr) to Z = 52 (Te) have been constructed in order to comment on migration/retention of these elements at Oklo. Although data for fissiogenic amounts of some of these elements are lacking, where such data are available to agreement between predicted migration/retention based on the Eh-pH diagrams and actual measurement is excellent. Based on Eh-pH diagrams, migration (to what degree is uncertain) of Mo and Cd is predicted whereas retention of Y, Zr, Tc, Ru, Rh, Pd, Ag and Te is also predicted. An earlier report of Frejacques et al. of Tc migration is in disagreement with Eh-pH prediction, and recent (unpublished) data argue for Tc retention. In view of the agreement between prediction and observation, the possible migration of Sb and retention of In and Sn is proposed. These data again demonstrate the usefulness of Eh-pH diagrams for the Oklo fossil nuclear reactor but, more important, allow constraints to be placed on repositories for nuclear waste now under consideration.

Issues of high-burnup fuel for advanced nuclear reactors

International Nuclear Information System (INIS)

Belac, J.; Milisdoerfer, L.

2004-12-01

A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

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

Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

type of reactors: low power reactors, graphite and natural uranium reactors, heavy water and natural uranium reactors (with water, gas or heavy water as coolant), heavy water and enriched uranium reactors, 'kettle' type reactors, pool type reactors and high power reactors. Finally, additional factors are considered as the correlation of the reactor type with the research program, safety considerations, size of the construction and economical considerations. (M.P.)

Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

type of reactors: low power reactors, graphite and natural uranium reactors, heavy water and natural uranium reactors (with water, gas or heavy water as coolant), heavy water and enriched uranium reactors, 'kettle' type reactors, pool type reactors and high power reactors. Finally, additional factors are considered as the correlation of the reactor type with the research program, safety considerations, size of the construction and economical considerations. (M.P.)

Polarization of fast neutrons in VVR-M reactor

International Nuclear Information System (INIS)

Garusov, E.A.; Lifshits, E.P.; Petrov, Yu.V.

1987-01-01

Neutron polarization in the reactor leads to circular polarization of γ quanta emitted both in radiational capture of neutrons and in the transition of nuclei excited as a result of inelastic scattering to the ground state. This may be used to determine the polarization of reactor neutrons. The circular polarization of γ quanta at light-water and graphite targets at the center of the active zone of the VVR-M reactor at the B.P. Konstantinov Leningrad Institute of Nuclear Physics was recently measured. A simplified experimental scheme is shown. Fast neutrons leaving the active zone of the reactor were excited in the inelastic scattering at the target nuclei. The polarization of the γ quanta emitted by nuclei in transitions to the ground state was measured by a polarimeter positioned above the active zone. The reason for the circular polarization of γ quanta may also be nonconservation of P parity on account of weak interaction in the capture of a neutron by hydrogen

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.)

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

Methods in nuclear reactors calculations; Metodos de calculo en reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Velarde, G

1966-07-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P{sub l}; B{sub l}; M{sub l}; S{sub n} and discrete ordinates approximations. (Author)

Safety assessment of Department of Energy nuclear reactors

International Nuclear Information System (INIS)

1981-03-01

One of the first tasks of the NFPQT Committee was to determine which DOE reactors would be assessed. The Committee determined that in view of the limited time available to conduct the assessment, 13 DOE reactors were of such size (physical, power or fission product inventory) to warrant review. This determination was approved by the Under Secretary. A decision was also made in the cases of three weapons material production reactors, C, K and P, to concentrate on the K reactor only, since all three are of the same basic design, have the same operating features, are all at the same site, and are all operated by the same contractor. The assessment was accomplished in the following ways: reviewing the results of assessments conducted by the DOE organizations with reactor safety responsibilities, which were undertaken in compliance with the request of the various program directors; reviewing selected documents that were requested by the Committee and assembled at DOE Headquarters; interviewing DOE Headquarters and Field Office personnel; and conducting on-site reviews of four reactors located at four different sites. The four reactors for on-site reviews were: Advanced Test Reactor (ATR); K Production Reactor; High Flux Beam Reactor (HFBR); and High Flux Isotope Reactor (HFIR). Specific findings and recommendations from the assessment are presented

Evaluation of Packed-Bed Reactor and Continuous Stirred Tank Reactor for the Production of Colchicine Derivatives

OpenAIRE

Dubey, Kashyap Kumar; Kumar, Dhirendra; Kumar, Punit; Haque, Shafiul; Jawed, Arshad

2013-01-01

Bioconversion of colchicine into its pharmacologically active derivative 3-demethylated colchicine (3-DMC) mediated by P450BM3 enzyme is an economic and promising strategy for the production of this inexpensive and potent anticancer drug. Continuous stirred tank reactor (CSTR) and packed-bed reactor (PBR) of 3 L and 2 L total volumes were compared for the production of 3-demethylated colchicine (3-DMC) a colchicine derivative using Bacillus megaterium MTCC*420 under aerobic conditions. Statis...

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

Activation analysis with small mobile reactors

International Nuclear Information System (INIS)

Chung, C.

1990-01-01

A small nuclear reactor (a low-power reactor without heat removal devices) usually has thermal power output under 100 W and an average in-core thermal neutron flux below 10 9 n/cm 2 s. Conventional activation analysis is restricted to determination of specific elements with large neutron capture cross sections in sizable samples. In-vivo prompt gamma activation analysis (IVPGAA) can be used for diagnosis of elemental composition of the human body, particularly the essential elements Ca, Cl, N, and P in the whole body, and toxic Cd and Hg in contaminated organs. In this chapter, activation analysis using an external neutron beam from the Tsing Hua Mobile Educational Reactor (THMER) for in vivo activation is described. Characteristics of the mobile reactor, in-vivo medical diagnosis, and radiation safety are emphasized. 17 refs, 12 figs, 3 tabs

Acclimatization of anaerobic sludge for UASB-reactor start-up

NARCIS (Netherlands)

Zeeuw, de W.J.

1984-01-01

<p/>The Upflow Anaerobic Sludge Bed (UASB) reactor represents a high rate anaerobic wastewater treatment system. The majority of the active biomass in the reactor is present in the form of sludge granules which possess excellent settling properties.
If no acclimatized (granular)

Anaerobic degradation of coconut husk leachate using UASB-reactor.

Science.gov (United States)

Neena, C; Ambily, P S; Jisha, M S

2007-07-01

Reffing of coconut husk, the majorprocess in quality coir fibre extraction, causes serious pollution with brackish water lagoons of Kerala. An attempt is made to treat the coconut husk leachate by using a laboratory scale UASB-reactor The experiment was conducted with loading of leachate from 1 kg of fresh coconut husk. The anaerobic treatment was done continuously The parameters like VFA, pH, COD and polyphenols were analysed regularly during the evaluation of the reactor performance. The polyphenol, VFA and COD were diminished gradually with time. The pH of the reactor during the study was found to be in the range of 6-8. The biogas production was increased with loading and about 82% of the total COD/kg husk could be converted to biogas. The maximum polyphenol loading in the reactor was reached to about 298.51 mg/l of husk.

Homogeneous Thorium Fuel Cycles in Candu Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

2009-06-15

The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

On the optimization of a steady-state bootstrap-reactor

International Nuclear Information System (INIS)

Polevoy, A.R.; Martynov, A.A.; Medvedev, S.Yu.

1993-01-01

A commercial fusion tokamak-reactor may be economically acceptable only for low recirculating power fraction r 0 ≡ P CD /P α BS ≡I BS /I > 0.9 to sustain the steady-state operation mode for high plasma densities > 1.5 10 20 m -3 , fulfilled the divertor conditions. This paper presents the approximate expressions for the optimal set of reactor parameters for r BS /I∼1, based on the self-consistent plasma simulations by 1.5D ASTRA code. The linear MHD stability analysis for ideal n=1 kink and ballooning modes has been carried out to determine the conditions of stabilization for bootstrap steady state tokamak reactor BSSTR configurations. (author) 10 refs., 1 tab

Radiation detectors for reactors

International Nuclear Information System (INIS)

Balagi, V.

2005-01-01

Detection and measurement of radiation plays a vital role in nuclear reactors from the point of view of control and safety, personnel protection and process control applications. Various types of radiation are measured over a wide range of intensity. Consequently a variety of detectors find use in nuclear reactors. Some of these devices have been developed in Electronics Division. They include gas-filled detectors such as 10 B-lined proportional counters and chambers, fission detectors and BF 3 counters are used for the measurement of neutron flux both for reactor control and safety, process control as well as health physics instrumentation. In-core neutron flux instrumentation employs the use detectors such as miniature fission detectors and self-powered detectors. In this development effort, several indigenous materials, technologies and innovations have been employed to suit the specific requirement of nuclear reactor applications. This has particular significance in view of the fact that several new types of reactors such as P-4, PWR and AHWR critical facilities, FBTR, PFBR as well as the refurbishment of old units like CIRUS are being developed. The development work has sought to overcome some difficulties associated with the non-availability of isotopically enriched neutron-sensing materials, achieving all-welded construction etc. The present paper describes some of these innovations and performance results. (author)

Reactors. Nuclear propulsion ships

International Nuclear Information System (INIS)

Fribourg, Ch.

2001-01-01

This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

International Nuclear Information System (INIS)

Febrianto

2006-01-01

Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

Design of first reactor protection system prototype for C A R E M reactor

International Nuclear Information System (INIS)

Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

2006-01-01

In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

Hybrid nuclear reactors and muon catalysis

International Nuclear Information System (INIS)

Petrov, Yu.

1983-01-01

Three methods are described of the conversion of isotope 238 U to 239 Pu by neutron capture in fast breeder reactors, in the breeding blanket of hybrid thermonuclear reactors using neutrons generated by fusion and electronuclear breeding in which the target is bombarded with 1 GeV protons. Their possible use in power production is discussed. Another prospective energy source is the use of muon catalysis in the fusion of deuterium and tritium nuclei. (J.P.)

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.)

Performance and application of a fluidized bed limestone reactor designed for control of alkalinity, hardness and pH at the Warm Springs Regional Fisheries Center

Science.gov (United States)

Watten, Barnaby J.; Mudrak, Vincent A.; Echevarria, Carlos; Sibrell, Philip; Summerfelt, Steven T.; Boyd, Claude E.

2017-01-01

Springs serving the Warm Springs Regional Fisheries Center, Warm Springs, Georgia, have pH, alkalinity, and hardness levels thatlie under the range required for successful fish propagation while free CO2 is well above allowable targets. We evaluate a pretreatment process that exploits limestone’s (CaCO3) ability to react away hydrogen ions (H+) and carbon dioxide (CO2) while increasing alkalinity (HCO3−) and calcium (Ca2+) concentrations, i.e. CaCO3 + H+ ↔ HCO3− + Ca2+ CaCO3 + CO2 + H2O ↔ Ca2+ + 2HCO3− Limestone sand was tested in both pilot and full scale fluidized bed reactors (CycloBio®). We first established the bed expansion characteristics of three commercial limestone products then evaluated the effect of hydraulic flux and bed height on dissolution rate of a single selected product (Type A16 × 120). Pilot scale testing at 18C showed limestone dissolution rates were relatively insensitive to flux over the range 1.51–3.03 m3/min/m2 but were sensitive (P changes in bed height (BH, cm) over the range 83–165 cm following the relation: (Alkalinity, mg/L) = 123.51 − (3788.76 (BH)). Differences between filtered and non-filtered alkalinity were small(P > 0.05) demonstrating that limestone was present in the reactor effluent primarily in the form of dissolved Ca(HCO3)2. Effluent alkalinity exceeded our target level of 50 mg/L under most operating conditions evaluated with typical pilot scale values falling within the range of 90–100 mg/L despite influent concentrations of about 4 mg/L. Concurrently, CO2 fell from an average of 50.6 mg/L to 8.3 mg/L (90%), providing for an increase in pH from 5.27 to a mean of 7.71. The ability of the test reactor to provide changes in water chemistry variables that exceeded required changes allowed for a dilution ratio of 0.6. Here, alkalinity still exceeded 50 mg/L, the CO2 concentration remained well below our limit of 20 mg/L (15.4 mg/L) and the pH was near neutral (7.17). Applying the dilution ratio of 0

Feasibility of biohydrogen production from cheese whey using a UASB reactor: Links between microbial community and reactor performance

Energy Technology Data Exchange (ETDEWEB)

Castello, E.; Garcia y Santos, C.; Borzacconi, L. [Chemical Engineering Institute, School of Engineering, University of the Republic, Herrera y Reissig 565, Montevideo (Uruguay); Iglesias, T.; Paolino, G.; Wenzel, J.; Etchebehere, C. [Microbiology Department, School of Science and School of Chemistry, University of the Republic, General Flores 2124, Montevideo (Uruguay)

2009-07-15

The present study examines the feasibility of producing hydrogen by dark fermentation using unsterilised cheese whey in a UASB reactor. A lab-scale UASB reactor was operated for more than 250 days and unsterilised whey was used as the feed. The evolution of the microbial community was studied during reactor operation using molecular biology tools (T-RFLP, 16S rRNA cloning library and FISH) and conventional microbiological techniques. The results showed that hydrogen can be produced but in low amounts. For the highest loading rate tested (20 gCOD/L.d), hydrogen production was 122 mL H{sub 2}/L.d. Maintenance of low pH (mean = 5) was insufficient to control methanogenesis; methane was produced concomitantly with hydrogen, suggesting that the methanogenic biomass adapted to the low pH conditions. Increasing the loading rate to values of 2.5 gCOD/gVSS.d favoured hydrogen production in the reactor. Microbiological studies showed the prevalence of fermentative organisms from the genera Megasphaera, Anaerotruncus, Pectinatus and Lactobacillus, which may be responsible for hydrogen production. However, the persistence of methanogenesis and the presence of other fermenters, not clearly recognised as hydrogen producers indicates that competition for the substrate may explain the low hydrogen production. (author)

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

Safety aspects of water chemistry in light water reactors

International Nuclear Information System (INIS)

1988-12-01

The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

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)

Fast neutron flux in the RA reactor experimental channels; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Dobrosavljevic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

Fast neutron flux in the RA reactor experimental channels was determined by using threshold reaction detectors. The (n,p) type reactions S{sub 32} (n,p)P{sub 32}, and Al{sub 24} (n,p)Na{sub 24}. Prepared sulphur and phosphorous foils were placed in cadmium boxes and irradiated in experimental channels VK-5, VK-7 and VK-9. Gold foils were irradiated simultaneously for controlling the reactor power. Reactor power was 100 kW during irradiation of half an hour. Activity of P{sub 32} and S{sub 31} after reactor shutdown was measured by 4{pi} counter and three calibrated GM counters. Absolute neutron flux was determined by using thus obtained data.

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)

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Fast reactor fuel reprocessing in the UK

International Nuclear Information System (INIS)

Allardice, R.H.; Williams, J.; Buck, C.

1977-01-01

Enriched uranium metal fuel irradiated in the Dounreay Fast Reactor has been reprocessed and refabricated in plants specifically designed for the purpose in the U.K. since 1961. Efficient and reliable fuel recycle is essential to the development of a plutonium based fast reactor system and the importance of establishing at an early stage fast reactor fuel reprocessing has been reinforced by current world difficulties in reprocessing high burn-up thermal reactor oxide fuel. In consequence, the U.K. has decided to reprocess irradiated fuel from the 250 MW(E) Prototype Fast Reactor as an integral part of the fast reactor development programme. Flowsheet and equipment development work for the small scale fully active demonstration plant have been carried out over the past 5 years and the plant will be commissioned and ready for active operation during 1977. In parallel, a comprehensive waste management system has been developed and installed. Based on this development work and the information which will arise from active operation of the plant a parallel development programme has been initiated to provide the basis for the design of a large scale fast reactor fuel reprocessing plant to come into operation in the late 1980s to support the projected U.K. fast reactor installation programme. The paper identifies the important differences between fast reactor and thermal reactor fuel reprocessing technologies and describes some of the development work carried out in these areas for the small scale P.F.R. fuel reprocessing operation. In addition, the development programme in aid of the design of a larger scale fast reactor fuel reprocessing plant is outlined and the current design philosophy is discussed

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Damage analysis of TRIGA MARK II Bandung reactor tank material structure

International Nuclear Information System (INIS)

Soedardjo; Sumijanto

2000-01-01

Damage of Triga Mark II Bandung reactor tank material structure has been analyzed. The analysis carried out was based on ultrasonic inspection result in 1996 and the monthly reports of reactor operation by random data during 1988 up to 1995. Ultrasonic test data had shown that thinning processes on south and west region of reactor out side wall at upper part of water level had happened. Reactor operation data had shown the demineralized water should be added monthly to the reactor and bulk shielding water tank. Both reactor and bulk shielding tank are shielded by concrete of Portland type I cement consisting of CaO content about 58-68 %. The analysis result shows that the reaction between CaO and seepage water from bulk shielding wall had taken place and consequently the reactor out sidewall surroundings became alkaline. Based on Pourbaix diagram, the aluminum reactor tank made of aluminum alloy 6061 T6 would be corroded easily at pH equal an greater than 8.6. The passive layer AI 2 O 3 aluminum metal surface would be broken due to water reaction taken place continuously at high pH and produces hydrogen gas. The light hydrogen gas would expand the concrete cement and its expanding power would open the passive layer of aluminum metal upper tank. The water sea pages from adding water into reactor tank could indicate the upper water level tank corrosion is worse than the lower water level tank. (author)

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)

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

Proceedings of the sixth Asian symposium on research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

Proceedings of the sixth Asian symposium on research reactors

International Nuclear Information System (INIS)

1999-08-01

The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: 1) status and future plan of research and testing reactors, 2) operating experiences, 3) design and modification of the facility, and reactor fuels, 4) irradiation studies, 5) irradiation facilities, 6) reactor characteristics and instrumentation, and 7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

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

Transmutation of americium in critical reactors

International Nuclear Information System (INIS)

Wallenius, J.

2005-01-01

Already in 1974, a Los Alamos report suggested that the recycling of higher actinides would be detrimental for the safety of critical reactors. Later investigations confirmed this understanding, and stringent limits on the fraction of minor actinides allowed to be present in the fuel of fast neutron reactors were established. In recent years, and in particular in connection with the generation IV initiative, it has been advocated that recycling of americium in critical reactors is not only feasible, but also a recommendable approach. In the present contribution, it is shown, to the contrary, that introduction of americium into reactors with uranium based fuels deteriorates the safety margin of these reactors to a degree that will not allow consumption of the americium sources present in any economically competitive nuclear fuel cycle. Further, it is shown that uranium and thorium free cores with plutonium based fuels may be designed, that features excellent safety characteristics, as long as americium is not present in the feed. Hence, a closed fuel cycle is suggested, that consists of commercial power production in light water reactors, plutonium burning in uranium and thorium free fast neutron critical reactors, and higher actinide consumption in accelerator driven systems with inert matrix fuel. It is argued that such a fuel cycle (being a refinement of the Double Strata fuel cycle proposed by JAERI and further developed by M. Salvatores) provides a minimum cost penalty for implementing P and T under realistic boundary conditions. (author)

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.)

Basis for Interim Operation for the K-Reactor in Cold Standby

Energy Technology Data Exchange (ETDEWEB)

Shedrow, B.

1998-10-19

The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993).

Basis for Interim Operation for the K-Reactor in Cold Standby

International Nuclear Information System (INIS)

Shedrow, B.

1998-01-01

The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993)

Hybrid biogas upgrading in a two-stage thermophilic reactor

DEFF Research Database (Denmark)

Corbellini, Viola; Kougias, Panagiotis; Treu, Laura

2018-01-01

The aim of this study is to propose a hybrid biogas upgrading configuration composed of two-stage thermophilic reactors. Hydrogen is directly injected in the first stage reactor. The output gas from the first reactor (in-situ biogas upgrade) is subsequently transferred to a second upflow reactor...... (ex-situ upgrade), in which enriched hydrogenotrophic culture is responsible for the hydrogenation of carbon dioxide to methane. The overall objective of the work was to perform an initial methane enrichment in the in-situ reactor, avoiding deterioration of the process due to elevated pH levels......, and subsequently, to complete the biogas upgrading process in the ex-situ chamber. The methane content in the first stage reactor reached on average 87% and the corresponding value in the second stage was 91%, with a maximum of 95%. A remarkable accumulation of volatile fatty acids was observed in the first...

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.

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

Evaluation of an automated struvite reactor to recover phosphorus ...

African Journals Online (AJOL)

In the present study we attempted to develop a reactor system to recover phosphorus by struvite precipitation, and which can be installed anywhere in the field without access to a laboratory. A reactor was developed that can run fully automated and recover up to 93% of total phosphorus (total P). Turbidity and conductivity ...

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

Study of Xenon-poisoning effect on the research reactor power

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.

2000-01-01

The uranium 235 is often used as a fuel to produce the energy in nuclear reactors. Uranium nuclei are fissioned with thermal neutrons and produce energy plus a number of neutrons. A fraction of such fission neutrons is involved in other fission with new nuclei to sustain the fission reactions. The remain fraction of the neutrons is lost from the reactor in two ways: escaped from the reactor, or absorbed with other nuclei that exist in the reactor before or produced from fission. Fission nuclei which absorb neutrons heavily are called p oison , such as Xe 135. Because Xe 135 absorbs neutrons heavily, it reduces the number of neutrons in the reactor. Hence, Xe 135 is studied explicitly in the MNSR reactor, and calculation of its negative reactivity is presented in this research during the operation, equilibrium, and after the shutting down of the reactor. (author)

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)

On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

International Nuclear Information System (INIS)

1994-01-01

The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

LMFBR transducer performance in SLSF tests P1 and P2

International Nuclear Information System (INIS)

English, J.J.; Anderson, T.T.; Kuzay, T.M.; Wilson, R.E.; Pedersen, D.R.; Kaiser, W.C.; Klingler, W.B.

1977-01-01

The reliability and problem areas of sodium-immersed thermocouples, pressure transducers and flowmeters are presented for experiments P1 and P2 of the Sodium Loop Safety Facility (SLSF). The SLSF is a doubly-contained sodium loop situated in a core position of the Engineering Test Reactor at the Idaho National Engineering Laboratory

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

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Determination of the lowest critical power levels 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; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

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)

Serious accidents of PWR type reactors for power generation

International Nuclear Information System (INIS)

2008-12-01

This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

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

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

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.)

Performance Testing of Hydrodesulfurization Catalysts Using a Single-Pellet-String Reactor

NARCIS (Netherlands)

Moonen, Roel; Ras, Erik Jan; Harvey, Clare; Alles, Jeroen; Moulijn, J.A.

2017-01-01

<p>Small-scale parallel trickle-bed reactors were used to evaluate the performance of a commercial hydrodesulfurization catalyst under industrially relevant conditions. Catalyst extrudates were loaded as a single string in reactor tubes. It is demonstrated that product sulfur levels and densities

Methods in nuclear reactors calculations

International Nuclear Information System (INIS)

Velarde, G.

1966-01-01

Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and P l ; B l ; M l ; S n and discrete ordinates approximations. (Author)

Magnetohydrodynamic stability regimes for steady state and pulsed reactors

International Nuclear Information System (INIS)

Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

1994-01-01

A tokamak reactor will operate at the maximum value of β≡2μ 0 left angle p right angle /B 2 that is compatible with magnetohydrodynamic (MHD) stability. This value depends on the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near unity, I BS /I P ∼1, which constrains the product of the inverse aspect ratio and the plasma poloidal β to be near unity, arepsilonβ P ∼1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during ARIES I, II/IV, and III and PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements in the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies, is also discussed. ((orig.))

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.)

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

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

Development of regulatory policy for SMART-P

Energy Technology Data Exchange (ETDEWEB)

Lee, S. H.; Lee, Y. H.; Moo, Philip; Koh, B. J.; Son, M. K.; Han, G. H.; Kim, D. H. [Korea Association for Nuclear Technology, Daejeon (Korea, Republic of)

2004-06-15

KAERI promoted the construction of a research reactor, SMART-P, the reduced scale of SMART, with intent to demonstrate the safety and performance of SMART. According to this progress, the development of regulatory process for SMART-P became necessary. The establishment of regulatory policy, based on the current regulatory guidelines as well as technical aspect, became essential matters. Considering the on-going small and medium size reactors in near future, the selection of the appropriate measure in the existing regulatory process to SMART-P is very important. Thus the schematic study for the applicable licensing procedure and regulatory requirements suitable for SMART-P is required.

Development of regulatory policy for SMART-P

Energy Technology Data Exchange (ETDEWEB)

Lee, S. H.; Moon, S. H.; Lee, Y. H.; Son, M. K.; Han, K. H.; Kim, D. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

2003-06-15

KAERI promoted the construction of a research reactor, SMART-P, the reduced scale of SMART, with intent to demonstrate the safety and performance of SMART. According to this progress, the development of regulatory process for SMART-P became necessary. The establishment of regulatory policy, based on the current regulatory guidelines as well as technical aspect, became essential matters. Considering the on-going small and medium size reactors m near future, the selection of the appropriate measure in the existing regulatory process to SMART-P is very important. Thus the schematic study for the applicable licensing procedure and regulatory requirements suitable for SMART-P is required.

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

Prototype moving-ring reactor

International Nuclear Information System (INIS)

Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

1981-01-01

The objective of this work was to design a prototype fusion reactor based on fusion plasmas confined as ''Compact Toruses.' Six major criteria guided the prototype design. The prototype must: (1) produce net electricity decisively (P/sub net/ >70% of P/sub gross/), with P/sub net/ approximately 100 MW(e); (2) have small physical size (low project cost) but commercial plant; (3) have all features required of commerical plants; (4) avoid unreasonable extrapolation of technology; (5) minimize nuclear issues substantially, i.e. accident and waste issues of public concern, and (6) be modular (to permit repetitive fabrication of parts) and be maintainable with low occupational radiological exposures

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.)

TRAC-P validation test matrix. Revision 1.0

International Nuclear Information System (INIS)

Hughes, E.D.; Boyack, B.E.

1997-01-01

This document briefly describes the elements of the Nuclear Regulatory Commission's (NRC's) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC's software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort

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.

Study of the 99Mo production effect on some of research reactor core specification using DARE-P

International Nuclear Information System (INIS)

Khamis, I.; Ezzuddin, H.

2006-12-01

In this study, the major basis of Mo- production using irradiation method for targets of enriched uranium is presented. Factors such as neutron flux, time of irradiation, and enrichment ration have also been analyzed. It was concluded that Mo-production requires nuclear reactors having high neutron fluxes, and that raising the neutron flux results in increasing the Mo-activities especially at fluxes that are higher than 10 14 n.cm -2 .s -1 . Time of irradiation was found to have adverse i.e. positive and negative effects on Mo activity. Therefore, the optimum irradiation time should be determined. The Mo-production was modeled through a set of differential equation with respect to time with aim to study the most influencing factors on Mo production, especially the effects of neutron flux and time of irradiation. In addition, the mass of plutonium produced during the irradiation of low enriched uranium has been evaluated. An investigation as to whether 99 Mo could be produced in the Syrian MNSR has been made. The result shows that a specific activity of 0.3773 Ci/g for 99 Mo could be produced. In future, a nuclear research reactor having neutron flux of 10 14 n.cm -2 .s -1 and 8 irradiation sites suitable for targets of 5 g of 235 U could produce about 5000 Ci on a weekly. Calculation and simulation have been achieved using the already developed DARE-P simulation language which was modified and improved in the nuclear Engineering Department in the Atomic Energy Commission of Syria. (author)

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.)

Application of the system engineering approach for reactor plants design

International Nuclear Information System (INIS)

Sitskiy, S. B.

2010-01-01

The main activities planned for to be implemented are: developing a data model of the reactor plant plus integration with the information model of the plant (3D model + P & ID); reengineering of processes, developing of electronic documents; description of the equipment for information management of the reactor plant lifecycle – according ISO15926

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.

Design Engineering of Safety Parameter Display for Kartini Reactor

International Nuclear Information System (INIS)

Yoyok Dwi Setyo Pambudi; Suharyo Widagdo; Aliq Zuhdi; Darlis

2003-01-01

Modification information parameter of Reactor Kartini has been conducted. This program use compiler LabView 5.0 that compatible and relevant with National Instrument card. These card mostly use in reactor instrumentation and industry. The information that been display is control rod position, fuel temperature, and water pH. The development in this research is control rod figure that can move up and down as the output from reactor power. Beside this is a module-module security figure that can light if there is a warning signal. Maximum high of control rod is 100 cm. (author)

Spatial distribution of reactor radiation around the horizontal experimental hole of the RA Reactor at Vinca; Prostorna raspodela reaktorskog zracenja oko horizontalnog eksperimentalnog kanala reaktora RA u Vinci

Energy Technology Data Exchange (ETDEWEB)

Ninkovic, M; Paligoric, D; Vujisic, B [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1964-10-15

A survey is given of the measurement of the spatial distribution of thermal and fast neutrons and gamma radiations inside and around the reactor radiation beam which is carried through the experimental channel outside the reactor biological shield thus being a potential danger for the personnel in the reactor hall. The results obtained are used to determine the optimal distribution of the shielding elements around the free radiation beam. The activation method was used to determine the distribution of slow and fast neutrons. The detectors were Au{sup 197} and In {sup 115} for slow, and the S{sup 32} (n,p) P{sup 32} reaction for fast neutrons. The relative distribution of slow and fast neutrons along the beam axis in the space outside the reactor, and evaluation of the absolute value of the flux at the place where the beam comes out from the biological shield are given (author)

A view of technology maturity assessment to realize fusion reactor by Japanese young researchers

International Nuclear Information System (INIS)

Kasada, Ryuta; Goto, Takuya; Miyazawa, Junichi; Fujioka, Shinsuke; Hiwatari, Ryoji; Oyama, Naoyuki; Tanigawa, Hiroyasu

2013-01-01

Japanese young researchers who have interest in realizing fusion reactor have analyzed Technology Readiness Levels (TRL) in Young Scientists Special Interest Group on Fusion Reactor Realization. In this report, brief introduction to TRL assessment and a view of TRL assessment against fusion reactor projects conducting in Japan. (J.P.N.)

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.

Kinetics of anaerobic digestion of labaneh whey in a batch reactor

African Journals Online (AJOL)

SAM

2014-04-16

Apr 16, 2014 ... kinetic constants were determined for labaneh whey and for diluted whey .... reactor has a pH and temperature control system. ... Variable power electric heater was used to heat the reactor. ..... by gas chromatography, Annual book of ASTM Standard, Vol. ... Thesis, The University of Jordan, Amman, Jordan.

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

TRAC-P validation test matrix. Revision 1.0

Energy Technology Data Exchange (ETDEWEB)

Hughes, E.D.; Boyack, B.E.

1997-09-05

This document briefly describes the elements of the Nuclear Regulatory Commission`s (NRC`s) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC`s software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort.

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.)

Proceedings of the specialists' meeting on reactor group constants

Energy Technology Data Exchange (ETDEWEB)

Katakura, Jun-ichi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

This report is the Proceedings of the Specialists' Meeting on Reactor Group Constants. The meeting was held on February 22-23, 2001 at Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of 59 specialists. The evaluation work for JENDL-3.3 is going on for the publication in a short time. The processing JENDL-3.3 file to make reactor group constants is needed when it is used in application fields. In the meeting, the present status of the reactor group constants was reviewed and the issues relating to them were discussed in such fields as thermal reactor, criticality safety, fast reactor, high energy region, burn-up calculation and radiation shielding. At the final session in the meeting, standardization of reactor group constants was discussed and the need of the reference group constants was confirmed by the participants. The 11 of the presented papers are indexed individually. (J.P.N.)

Decreasing methane production in hydrogenogenic UASB reactors fed with cheese whey

International Nuclear Information System (INIS)

Carrillo-Reyes, Julián; Celis, Lourdes B.; Alatriste-Mondragón, Felipe; Razo-Flores, Elías

2014-01-01

One of the problems in fermentative hydrogen producing reactors, inoculated with pre-treated anaerobic granular sludge, is the eventual methane production by hydrogen-consuming methanogens. In this study, strategies such as reduction of pH and HRT, organic shock loads and repeated biomass heat treatment were applied to hydrogenogenic UASB reactors fed with cheese whey, that showed methane production after certain time of continuous operation (between 10 and 60 days). The reduction of pH to 4.5 not only decreased methane production but also hydrogen production. Organic shock load (from 20 to 30 g COD/L-d) was the more effective strategy to decrease the methane production rate (75%) and to increase the hydrogen production rate (172%), without stopping reactor operation. Repeated heat treatment of the granular sludge was the only strategy that inhibited completely methane production, leading to high volumetric hydrogen production rates (1.67 L H 2 /L-d), however this strategy required stopping reactor operation; in addition homoacetogenesis, another hydrogen-consuming pathway, was not completely inhibited. This work demonstrated that it was possible to control the methane activity in hydrogen producing reactors using operational strategies. - Highlights: • Operational strategies control methane in hydrogen production from cheese whey. • Organic shock load increased the hydrogen production rate. • Operation pH below 5 decreased both the hydrogen and methane production. • Second biomass heat treatment inhibits completely methanogenesis. • Homoacetogens play a negative role in fermentative hydrogen production

Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

International Nuclear Information System (INIS)

Hirota, Jitsuya

1984-09-01

The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

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

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

2002-01-01

The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

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

The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

International Nuclear Information System (INIS)

Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

2010-01-01

LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

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

Experimental study and nuclear model calculations on the $^{192}Os (p, n)^{192}$Ir reaction Comparison of reactor and cyclotron production of the therapeutic radionuclide $^{192}$Ir

CERN Document Server

Hilgers, K; Sudar, S; 10.1016/j.apradiso.2004.12.010

2005-01-01

In a search for an alternative route of production of the important therapeutic radionuclide /sup 192/Ir (T/sub 1/2/=78.83 d), the excitation function of the reaction /sup 192/Os(p, n)/sup 192/Ir was investigated from its threshold up to 20MeV. Thin samples of enriched /sup 192/Os were obtained by electrodeposition on Ni, and the conventional stacked-foil technique was used for cross section measurements. The experimental data were compared with the results of theoretical calculations using the codes EMPIRE-II and ALICE-IPPE. Good agreement was found with EMPIRE-II, but slightly less with the ALICE-IPPE calculations. The theoretical thick target yield of /sup 192/Ir over the energy range E/sub p/=16 to 8MeV amounts to only 0.16MBq/ mu A.h. A comparison of the reactor and cyclotron production methods is given. In terms of yield and radionuclidic purity of /sup 192/Ir the reactor method appears to be superior; the only advantage of the cyclotron method could be the higher specific activity of the product.

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.

Proceedings of the technical committee on high conversion and high burnup reactors

International Nuclear Information System (INIS)

Shiroya, Seiji; Kanda, Keiji; Sekiya, Tamotsu

1990-02-01

The present issue is the proceedings of 'the Technical Committee on High Conversion and High Burnup Reactors' held at Kyoto University Research Reactor Institute (KURRI) on December 12 and 22, 1988. In this committee, members so much concerned with this theme were asked to report their recent accomplishment and activities. By such a program, the committee was intended to make a survey of future direction of research in this type of reactor. (J.P.N.)

Physics experiment on the Dragon reactor

Energy Technology Data Exchange (ETDEWEB)

Hunt, C.

1974-10-15

The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

Proceedings of the international symposium on materials testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kawamura, Hiroshi

2009-01-01

This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

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

Simultaneous hydrogen utilization and in situ biogas upgrading in an anaerobic reactor

DEFF Research Database (Denmark)

Luo, Gang; Johansson, Sara; Boe, Kanokwan

2012-01-01

. The methane production rate of the reactor with H2 addition was 22% higher, compared to the control reactor only fed with manure. The CO2 content in the produced biogas was only 15%, while it was 38% in the control reactor. However, the addition of hydrogen resulted in increase of pH (from 8.0 to 8.3) due......The possibility of converting hydrogen to methane and simultaneous upgrading of biogas was investigated in both batch tests and fully mixed biogas reactor, simultaneously fed with manure and hydrogen. Batch experiments showed that hydrogen could be converted to methane by hydrogenotrophic...

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)

Research on physical and chemical parameters of coolant in Light-Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

Selective removal of heavy metals from metal-bearing wastewater in a cascade line reactor.

Science.gov (United States)

Pavlović, Jelena; Stopić, Srećko; Friedrich, Bernd; Kamberović, Zeljko

2007-11-01

This paper is a part of the research work on 'Integrated treatment of industrial wastes towards prevention of regional water resources contamination - INTREAT' the project. It addresses the environmental pollution problems associated with solid and liquid waste/effluents produced by sulfide ore mining and metallurgical activities in the Copper Mining and Smelting Complex Bor (RTB-BOR), Serbia. However, since the minimum solubility for the different metals usually found in the polluted water occurs at different pH values and the hydroxide precipitates are amphoteric in nature, selective removal of mixed metals could be achieved as the multiple stage precipitation. For this reason, acid mine water had to be treated in multiple stages in a continuous precipitation system-cascade line reactor. All experiments were performed using synthetic metal-bearing effluent with chemical a composition similar to the effluent from open pit, Copper Mining and Smelting Complex Bor (RTB-BOR). That effluent is characterized by low pH (1.78) due to the content of sulfuric acid and heavy metals, such as Cu, Fe, Ni, Mn, Zn with concentrations of 76.680, 26.130, 0.113, 11.490, 1.020 mg/dm3, respectively. The cascade line reactor is equipped with the following components: for feeding of effluents, for injection of the precipitation agent, for pH measurements and control, and for removal of the process gases. The precipitation agent was 1M NaOH. In each of the three reactors, a changing of pH and temperature was observed. In order to verify. efficiency of heavy metals removal, chemical analyses of samples taken at different pH was done using AES-ICP. Consumption of NaOH in reactors was 370 cm3, 40 cm3 and 80 cm3, respectively. Total time of the experiment was 4 h including feeding of the first reactor. The time necessary to achieve the defined pH value was 25 min for the first reactor and 13 min for both second and third reactors. Taking into account the complete process in the cascade line

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.)

Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

International Nuclear Information System (INIS)

Hussein, H.M.; Sakr, A.M.; Amin, E.H.

2011-01-01

Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

Reactor neutron activation analysis for aluminium in the presence of phosphorus and silicon. Contributions of /sup 28/Al activities from /sup 31/P (n,. cap alpha. ) /sup 28/Al and /sup 28/Si (n,p) /sup 28/Al reactions

Energy Technology Data Exchange (ETDEWEB)

Mizumoto, Yoshihiko (Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology); Iwata, Shiro; Sasajima, Kazuhisa; Yoshimasu, Fumio; Yase, Yoshiro

1984-01-01

Reactor neutron activation analysis for aluminium in samples containing phosphorus and silicon was studied. The experiments were performed by using pneumatic tube of the Kyoto University Reactor (KUR). At first, the ratios of the /sup 28/Al activity produced from /sup 27/Al(n, ..gamma..) /sup 28/Al reaction by thermal neutrons to that from /sup 31/P(n, ..cap alpha..)/sup 28/Al reaction by fast neutrons, and to that from /sup 28/Si(n, p)/sup 28/Al reaction were measured by ..gamma..-ray spectrometry. With a ratio of about 5 for the thermal to fast neutron flux of KUR, the ratio of the /sup 28/Al activity from aluminium to that from phosphorus was to be 812 +- 7, and to that from silicon 282 +- 3. Secondly, the contributions of /sup 28/Al activities from phosphorus and silicon and the determination limit of aluminium were calculated for various parameters, such as fast neutron flux, thermal to fast neutron flux ratio, amounts of phosphorus and silicon, etc. Thirdly, on the basis of these results, aluminium contents in spinal cords and brains of amyotrophic lateral sclerosis, Parkinsonism-dementia complex and control cases were determined.

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

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

should make them attractive for remote areas or electrical grids that are not large enough to support a standard-sized nuclear power plant. The last application of fast reactors this thesis investigates promises a solution to the problem of the radioactive waste from nuclear energy production. The separation of the spent fuel in different material streams (partitioning) and the irradiation of minor actinides in a fast neutron spectrum (transmutation) is claimed to solve this problem. Implementation of partitioning and transmutation (P and T) would require centuries of dedicated efforts, since several irradiation cycles and repeated reprocessing of the spent fuel elements between the irradiation cycles would be necessary. For all three applications, computer models of exemplary reactor systems were set up to perform criticality, depletion, and dose rate calculations. Based on the results, a specific critique on the viability of these fast reactor applications was conducted. Possible risks associated with their deployment were investigated.

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

should make them attractive for remote areas or electrical grids that are not large enough to support a standard-sized nuclear power plant. The last application of fast reactors this thesis investigates promises a solution to the problem of the radioactive waste from nuclear energy production. The separation of the spent fuel in different material streams (partitioning) and the irradiation of minor actinides in a fast neutron spectrum (transmutation) is claimed to solve this problem. Implementation of partitioning and transmutation (P and T) would require centuries of dedicated efforts, since several irradiation cycles and repeated reprocessing of the spent fuel elements between the irradiation cycles would be necessary. For all three applications, computer models of exemplary reactor systems were set up to perform criticality, depletion, and dose rate calculations. Based on the results, a specific critique on the viability of these fast reactor applications was conducted. Possible risks associated with their deployment were investigated.

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.)

Performance evaluation of full scale UASB reactor in treating stillage wastewater

Directory of Open Access Journals (Sweden)

A.Mirsepasi , H. R. Honary , A. R. Mesdaghinia, A. H. Mahvi , H. Vahid , H. Karyab

2006-04-01

Full Text Available Upflow anaerobic sludge blanket (UASB reactors have been widely used for treatment of industrial wastewater. In this study two full-scale UASB reactors were investigated. Volume of each reactor was 420 m3. Conventional parameters such as pH, temperature and efficiency of COD, BOD, TOC removal in each reactor were investigated. Also several initial parameters in designing and operating of UASB reactors, such as upflow velocity, organic loading rate (OLR and hydraulic retention time were investigated. After modifying in operation conditions in UASB-2 reactor, average COD removal efficiency at OLR of 10–11 kg COD / m3 day was 55 percent. In order to prevent solids from settling, upflow velocity was increased to 0.35 m/h. Also to prevent solids from settling, the hydraulic retention time of wastewater in UASB-2 reactor was increased from 200 to 20 hours. This was expected that with good operation of UASB-2 reactor and with expanding of granules in the bed of the reactor, COD removal efficiency will be increased to more than 80 percent. But, because of deficiency on granulation and operation in UASB-2 reactor, this was not achieved. COD removal efficiency in the UASB-1 reactor was little. To enhance COD efficiency of UASB-1 reactor, several parameters were needed to be changed. These changes included enhancing of OLRs and upflow velocity, decreasing hydraulic retention time and operating with new sludge.

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

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

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.

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)

Butanol production by Clostridium acetobutylicum in a continuous packed bed reactor.

Science.gov (United States)

Napoli, Fabio; Olivieri, Giuseppe; Russo, Maria Elena; Marzocchella, Antonio; Salatino, Piero

2010-06-01

In this study, we report on a butanol production process by immobilized Clostridium acetobutylicum in a continuous packed bed reactor (PBR) using Tygon rings as a carrier. The medium was a solution of lactose (15-30 g/L) and yeast extract (3 g/L) to emulate the cheese whey, an abundant lactose-rich wastewater. The reactor was operated under controlled conditions with respect to the pH and to the dilution rate. The pH and the dilution rate ranged between 4 and 5, the dilution rate between 0.54 and 2.4 h(-1) (2.5 times the maximum specific growth rate assessed for suspended cells). The optimal performance of the reactor was recorded at a dilution rate of 0.97 h(-1): the butanol productivity was 4.4 g/Lh and the selectivity of solvent in butanol was 88%(w).

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

International Nuclear Information System (INIS)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying; Lee, Hyung-Sool; Wang, Ai-Jie

2012-01-01

Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m −3 d −1 ) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m −3 d −1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

Energy Technology Data Exchange (ETDEWEB)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

2012-11-15

Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

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.)

Application of the exact distribution pj{sub k} in the determination of kinetic parameters in a reactor; Aplicacion de la distribucion exacta p{sub k} a la determinacion de parametros cineticos de un reactor

Energy Technology Data Exchange (ETDEWEB)

Alca Ruiz, F

1982-07-01

In this report one distribution of neutron counts obtained by a detector placed in a reactor is studied in order to be used in the determination of reactor kinetic parameters such as {beta}/{lambda} and reactivities. The parameters accuracy from this new method is compared with the Feynman and Mogilner method, based too in Reactor Neutron Noise Analysis. These three methods have been applied to JEN-2 reactor and the better accuracy and faster collection of experimental data give some interest to the new method which only requires a good footing code. (Author) 68 refs.

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.)

Anaerobic biogranulation in a hybrid reactor treating phenolic waste

International Nuclear Information System (INIS)

Ramakrishnan, Anushyaa; Gupta, S.K.

2006-01-01

Granulation was examined in four similar anaerobic hybrid reactors 15.5 L volume (with an effective volume of 13.5 L) during the treatment of synthetic coal wastewater at the mesophilic temperature of 27 ± 5 deg. C. The hybrid reactors are a combination of UASB unit at the lower part and an anaerobic filter at the upper end. Synthetic wastewater with an average chemical oxygen demand (COD) of 2240 mg/L, phenolics concentration of 752 mg/L and a mixture of volatile fatty acids was fed to three hybrid reactors. The fourth reactor, control system, was fed with a wastewater containing sodium acetate and mineral nutrients. Coal waste water contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. A mixture of anaerobic digester sludge and partially granulated sludge (3:1) were used as seed materials for the start up of the reactors. Granules were observed after 45 days of operation of the systems. The granules ranged from 0.4 to 1.2 mm in diameter with good settling characteristics with an SVI of 12 mL/g SS. After granulation, the hybrid reactor performed steadily with phenolics and COD removal efficiencies of 93% and 88%, respectively at volumetric loading rate of 2.24 g COD/L d and hydraulic retention time of 24 h. The removal efficiencies for phenol and m/p-cresols reached 92% and 93% (corresponding to 450.8 and 153 mg/L), while o-cresol was degraded to 88% (corresponding to 51.04 mg/L). Dimethyl phenols could be removed completely at all the organic loadings and did not contribute much to the residual organics. Biodegradation of o-cresol was obtained in the hybrid-UASB reactors

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.)

Use Of Cementitious Materials For SRS Reactor Facility In-Situ Decommissioning - 11620

International Nuclear Information System (INIS)

Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

2010-01-01

The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

USE OF CEMENTITIOUS MATERIALS FOR SRS REACTOR FACILITY IN-SITU DECOMMISSIONING - 11620

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

2010-12-07

The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

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.)

Continuous hyperpolarization with parahydrogen in a membrane reactor

Science.gov (United States)

Lehmkuhl, Sören; Wiese, Martin; Schubert, Lukas; Held, Mathias; Küppers, Markus; Wessling, Matthias; Blümich, Bernhard

2018-06-01

Hyperpolarization methods entail a high potential to boost the sensitivity of NMR. Even though the "Signal Amplification by Reversible Exchange" (SABRE) approach uses para-enriched hydrogen, p-H2, to repeatedly achieve high polarization levels on target molecules without altering their chemical structure, such studies are often limited to batch experiments in NMR tubes. Alternatively, this work introduces a continuous flow setup including a membrane reactor for the p-H2, supply and consecutive detection in a 1 T NMR spectrometer. Two SABRE substrates pyridine and nicotinamide were hyperpolarized, and more than 1000-fold signal enhancement was found. Our strategy combines low-field NMR spectrometry and a membrane flow reactor. This enables precise control of the experimental conditions such as liquid and gas pressures, and volume flow for ensuring repeatable maximum polarization.

Calculations of radiation defect formation cross sections in reactor materials in (n,p) and (n,α) reactions

International Nuclear Information System (INIS)

Kupchishin, A.A.; Kupchishin, A.I.; Omarbekova, Zh.

2001-01-01

In the work an experimental data analysis by integral σ(E 1 ) and differential [dσ(E 1 ,E 2 )]/dE 2 neutron interaction cross sections with reactor materials with the secondary protons and alpha particles generation as well as with the primarily knock-on atoms production in such reactions are carried out. It is shown, that in the (n,p) and (n',α) reactions the recoil nuclei receive essential energy portion and they are the patriarchs for atom-atom cascades in the substance. Nuclear reactions with formation of the secondary α-particles and and recoil nuclei are considered. It is shown, that these reactions are effectively proceeding within neutrons energy range 0.3-15 MeV. The nuclear reactions kinematics of above mentioned processes is studied. Energy conservation law for these reaction is applied. Deferential cross section conservation and transformation law for radiation defect formation in the (n,α) reaction are considered as well

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

Microorganism selection and biosurfactant production in a continuously and periodically operated bioslurry reactor.

Science.gov (United States)

Cassidy, D P; Hudak, A J

2001-06-29

A continuous-flow reactor (CSTR) and a soil slurry-sequencing batch reactor (SS-SBR) were maintained in 8l vessels for 180 days to treat a soil contaminated with diesel fuel (DF). Concentrations of Candida tropicalis, Brevibacterium casei, Flavobacterium aquatile, Pseudomonas aeruginosa, and Pseudomonas fluorescens were determined using fatty acid methyl ester (FAME) analysis. DF removal (biological and volatile) and biosurfactant concentrations were measured. The SS-SBR encouraged the growth of biosurfactant-producing species relative to the CSTR. Counts of biosurfactant-producing species (C. tropicalis, P. aeruginosa, P. fluorescens) relative to total microbial counts were 88% in the SS-SBR and 23% in the CSTR. Biosurfactants were produced in the SS-SBR to levels of nearly 70 times the critical micelle concentration (CMC) early in the cycle, but were completely degraded by the end of each cycle. No biosurfactant production was observed in the CSTR. DF biodegradation rates were over 40% greater and DF stripping was over five times lower in the SS-SBR than the CSTR. However, considerable foaming occurred in the SS-SBR. Reversing the mode of operation in the reactors on day 80 caused a complete reversal in microbial consortia and reactor performance by day 120. These results show that bioslurry reactor operation can be manipulated to control overall reactor performance.

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

Envisioning Communications with Future Stakeholders - A Case Study Using the In-Situ Decommissioning of P-Reactor

International Nuclear Information System (INIS)

Antonucci, D.L.

2009-01-01

This paper will explore opportunities to expand the CAB's public outreach by the incorporation of technologies typically used in social networks and distance learning. Envisioning opportunities to engage next generation CAB members in public involvement will be delineated by retracing the decision process used with the in-situ decommissioning of P-Reactor at the Savannah River Site (SRS). This paper will discuss existing opportunities to enable another group of stakeholders to take part in the environmental policy decision making process regarding the inclusion of some very long lived radioactive constituents. The aim of the paper will be to locate places in the current process where alternate or parallel informational dissemination pathways could exist. These alternatives will incorporate the next generation's expectation for instantaneous information and universal ownership of hand-held communication devices. The goal of this paper is to use the present framework of CAB communications and add the components of virtual networking and distance learning in hopes of bridging the generational technology gap and extending the dialog to future stakeholders. (authors)

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.)

Heterogeneous photocatalytic degradation of p-toluenesulfonic acid using concentrated solar radiation in slurry photoreactor

International Nuclear Information System (INIS)

Kamble, Sanjay P.; Sawant, Sudhir B.; Pangarkar, Vishwas G.

2007-01-01

In this work, the photocatalytic degradation (PCD) of p-toluenesulfonic acid (p-TSA) in batch reactor using concentrated solar radiation was investigated. The effect of the various operating parameters such as initial concentration of substrate, catalyst loading, solution pH and types of ions on photocatalytic degradation has been studied in a batch reactor to derive the optimum conditions. The rate of photocatalytic degradation was found to be maximum at the self pH (pH 3.34) of p-TSA. It was also observed that in the presence of anions and cations, the rate of PCD decreases drastically. The kinetics of photocatalytic degradation of p-TSA was studied. The PCD of p-TSA was also carried at these optimized conditions in a bench scale slurry bubble column reactor using concentrated solar radiation

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

The dynamic pressure measurements of the nuclear reactor coolant for condition-based maintenance of the reactor

International Nuclear Information System (INIS)

Es-Saheb, M.H.H.

1990-01-01

The condition-based maintenance of the nuclear reactor, by monitoring and measuring the instantaneous dynamic pressure distribution of the coolant (water) impact on the solid surfaces of the reactor during operation is presented. The behaviour of water domes (jets) produced by underwater explosions of small changes of P.E.T.N. at various depths in two different size cylindrical containers, which simulate the nuclear reactor, is investigated. Water surface domes (jets) from the underwater explosions are photographed. Depending on the depth of the charge, curved and flat top jets of up to 455 mm diameter and impact speeds of up to 70 m/sec. are observed. The instabilities in the dome surfaces are observed and the instantaneous profiles are analysed. It is found that, in all cases tested, the maximum pressure takes place at the center of the jet and could reach up to 3.0 times the on-dimensional impact pressure value. The use of their measurements, as online monitoring for condition-based maintenance and design-out maintenance is discussed. 18 refs

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)

Identification of reactor failure states using noise methods, and spatial power distribution

International Nuclear Information System (INIS)

Vavrin, J.; Blazek, J.

1981-01-01

A survey is given of the results achieved. Methodical means and programs were developed for the control computer which may be used in noise diagnostics and in the control of reactor power distribution. Statistical methods of processing the noise components of the signals of measured variables were used for identifying failures of reactors. The method of the synthesis of the neutron flux was used for modelling and evaluating the reactor power distribution. For monitoring and controlling the power distribution a mathematical model of the reactor was constructed suitable for control computers. The uses of noise analysis methods are recommended and directions of further development shown. (J.P.)

Ventilation air conditioner for a reactor container

International Nuclear Information System (INIS)

Ikegame, Noboru; Nakagawa, Takeshi.

1980-01-01

Purpose: To suppress the variations in the internal pressure of a reactor container and smoothly ventilate the reactor container. Constitution: The air conditioner provides an air-flow-rate-control damper, a purge-air supply fan, and a filter device in the air-supply pipe of a reactor container. Furthermore, it provides a pressure difference detector at a part of the container. The air-flow-rate-control damper is connected electrically through a position-modulator-comparison amplifier to the pressure difference detector. When the filtration becomes insufficient by clogging of the filter device and the internal pressure increased abruptly in the container, the pressure-difference detector can detect it, and the damper is operated by a pressure regulator and the comparator so as to control the air flow to the container. Thus, the internal pressure variation is controlled so as to easily ventilate the container. (J.P.N.)

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

MHD stability regimes for steady state and pulsed reactors

International Nuclear Information System (INIS)

Jardin, S.C.; Kessel, C.E.; Pomphrey, N.

1994-02-01

A tokamak reactor will operate at the maximum value of β≡2μ 0 /B 2 that is compatible with MHD stability. This value depends upon the plasma current and pressure profiles, the plasma shape and aspect ratio, and the location of nearby conducting structures. In addition, a steady state reactor will minimize its external current drive requirements and thus achieve its maximum economic benefit with a bootstrap fraction near one, I bs /I p ∼ 1, which constrains the product of the inverse aspect ratio and the plasma poloidal beta to be near unity, ε β p ∼ 1. An inductively driven pulsed reactor has different constraints set by the steady-state Ohm's law which relates the plasma temperature and density profiles to the parallel current density. We present the results obtained during the ARIES I, II/IV, and III and the PULSAR reactor studies where these quantities were optimized subject to different design philosophies. The ARIES-II/IV and ARIES-III designs are both in the second stability regime, but differ in requirements on the form of the profiles at the plasma edge, and in the location of the conducting wall. The relation between these, as well as new attractive MHD regimes not utilized in the ARIES or PULSAR studies is also discussed

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

Use of Cementitious Materials for SRS Reactor Facility In-Situ Decommissioning

International Nuclear Information System (INIS)

Langton, C.A.; Stefanko, D.B.; Serrato, M.G.; Blankenship, J.K.; Griffin, W.G.; Long, J.T.

2013-01-01

The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD project requires approximately 250000 cubic yards of cementitious materials to fill the below-grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Funding is being provided under the American Recovery and Reinvestment Act (ARRA). Cementitious materials were designed for the following applications: (A) Below-grade massive voids / rooms: Portland cement-based structural flowable fills for: (A.1) Bulk filling; (A.2) Restricted placement and (A.3) Underwater placement. (B) Special below-grade applications for reduced load bearing capacity needs: (B.1) Cellular portland cement lightweight fill. (C) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels (C.1) Blended calcium aluminate - calcium sulfate based flowable fill; (C.2) Magnesium potassium phosphate flowable fill. (D) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: (D.1) Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

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

2004-01-01

in the regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

To question of NPP power reactor choice for Kazakhstan

International Nuclear Information System (INIS)

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

2004-01-01

regions of Republic of Kazakhstan occurs. Southern and western regions import electric power and capacity because of undeveloped circuit of networks. Moreover, power intensity of an industrial-agrarian complex of the country is limited transmission capacity of lines is insufficient; plenty of small consumers are removed from power supply lines. Thus, nuclear stations of medium and low power are the most acceptable for construction in Kazakhstan. Recommendations for the choice of maximum safe, reliable and economically competitive reactors for Kazakhstan have been made in result of the carried out projects' comparison of the power reactors according to 15 criteria of safety and economic competitiveness, with respect to condition and perspectives of Kazakhstan power complex development: Recommended power reactors of medium capacity: - P-600 - passive PWR, developed of the Westinghouse company, USA; - CANDU-6, developed by Atomic Energy of Canada, Limited (AECL), Canada. - MS-600 - Mitsubishi Company, Japan. Recommended reactors of low power: - IRIS - reactor of IV generation developed by the international corporation of 13 organizations from 7 countries; - NPP 'UNITERM' - development NIKIET, Moscow, Russia; - MRX - the Project of sea reactor MRX for civil applications, is developed by the Japanese Research Institute of Atomic Energy (JAERI). The most important advantages of recommended medium and low power reactors are given

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

A review of fast reactor programme in Japan

Energy Technology Data Exchange (ETDEWEB)

Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

1981-05-01

The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

A review of fast reactor programme in Japan

International Nuclear Information System (INIS)

Masuno, Y.; Bando, S.

1981-01-01

The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

Ageing investigation and upgrading of components/systems of Kartini research reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

1998-10-01

Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

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)

Fatigue and environmentally assisted cracking in light water reactors

International Nuclear Information System (INIS)

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

1991-12-01

Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

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

Limits and conditions for permanent operation of the LVR-15 reactor

International Nuclear Information System (INIS)

1995-03-01

The ''Limits and Conditions'' have been set up to demonstrate compliance with all nuclear safety requirements specified by the Decree of the Czechoslovak Atomic Energy Commission No. 9/1985, and to serve as a guide for the reactor operators to respond adequately to deviations from the required values in normal operation conditions. The publications is divided into the following chapters: (1) Safety limits of the LVR-15 reactor; (2) Adjusting the reactor protection and control system parameters; (3) Limiting conditions for normal operation modes; (4) Control requirements; and (5) reactor operation management. A document by the Institute for Research, Production, and Application of Radioisotopes specifying conditions for the acceptance of radioactive wastes to the central repository is appended. (P.A.). 7 tabs

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

Life management for a non replaceable structure: the reactor building

International Nuclear Information System (INIS)

Torres, V.; Francia, L.

1998-01-01

Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

International Nuclear Information System (INIS)

Adamec, P.

2000-12-01

Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

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

Les réacteurs à membranes : possibilités d'application dans l'industrie pétrolière et pétrochimique Membrane Reactors: Possibilities of Application in the Petroleum and Petrochemical Industry

Directory of Open Access Journals (Sweden)

Guy C.

2006-11-01

Full Text Available Cet article fait le point sur l'état de la recherche dans le domaine des réacteurs chimiques avec séparation par membrane intégrée et de leur applications dans le domaine du raffinage et de la pétrochimie. Trois applications potentiellement intéressantes sont identifiées et, pour chacune, les avantages de l'utilisation d'un réacteur à membrane sont discutés. Ce sont : la déshydrogénation du propane en propylène, la déshydrogénation d'un naphtène cyclohexanique et le vaporéformage du gaz naturel. Pour ces réactions, les membranes à base de palladium apparaissent les plus performantes compte tenu de leur tenue en température, de leur sélectivité et de leur perméabilité à l'hydrogène. Quelques éléments relatifs à leur développement sont présentés en conclusion. Recently, the use of membrane in reaction engineering has been more and more advocated. The selective separation of the products from the reaction mixture allows to achieve higher conversion or better selectivity or to operate under less severe conditions or with smaller units. This paper presents an update on the recent advances in the field of chemical membrane reactors and on their applications in refining and petrochemistry. Previous work. Most of the possible applications of membrane reactors in petroleum and petrochemical industry concern gaseous catalytic reactions. For this reason, gas permeation membranes are the primary component of membrane reactors. Gas permeation membranes present different types of physical structure : dense, microporous or asymmetric which is a combination of the two. Separating properties of dense membranes are function of the solubility and diffusivity of each gaseous component in the membrane material. For microporous membranes, they follow four mechanisms : Knudsen diffusion, surface diffusion, capillary condensation or molecular sieving. Although organic polymers are the common constituent of gas permeation membrane, their

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.

RESI-1 and RESI-2: pPrototypes of an information system on reactor safety

International Nuclear Information System (INIS)

Schultheiss, G.F.; Eglin, W.; Katz, F.W.; Krings, T.; Pee, A.; Schlechtendahl, E.G.

1975-04-01

To demonstrate by practical experience the feasibility of the information system elaborated in the 'Study of an Information System on Reactor Safety RESI' (KFK 1900), the prototype systems RESI-1 and RESI-2 were developed and tested in operation. The two systems have been considerably reduced both in extent and contents as compared to the information system described in the study. The RESI-1 prototype system is a paper version established for verification of all the individual functions before passing over to the computer-aided interactive version RESI-2. RESI-2 is based on the GOLEM system of Siemens. Both protoype systems have proved that the essential features: 1) documentation, 2) formulation of and answering to safety questions, which are relevant with respect to particular licensing cases, 3) formulation of safety questions related to individual reactor types can be managed satisfactorily. All the functions of information retrieval have been tested carefully over several months. Particularities of project development and of the methods elaborated are described in detail and presented in this report. (orig.) [de

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)

Simplified simulation of an experimental fast reactor plant

International Nuclear Information System (INIS)

Fujii, Masaaki; Fujita, Minoru.

1978-01-01

Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

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)

Influence of decontamination of the WWER-440 primary circuit equipment on pressure drop in the reactor

International Nuclear Information System (INIS)

Kritsky, V.; Rodionov, Y.; Beresina, I.

2003-01-01

Over 40 reactor cycles at four WWER-440 type reactors have been analyzed in order to explain the increase of the pressure drop under certain combination of conditions. It is shown that the staff radiation exposure and the dose rate at first circuit segments are inversely correlated with the value of the pressure drop at the reactor, which is connected with the mechanism of redistribution of deposits and radioactive nuclides between the reactor and the rest part of the circuit. The influence of pH T on the formation of the dose rate from equipment and the change of pressure drop in the reactor WWER-440 is studied. The optimal range of pH T values for these parameters is determined to be 6.95-7.05 and these values are within the range of the water chemistry standards. The correlation between the changes of pressure drop and the number of decontaminated steam generators is established. This correlation shows that the pressure drop at the reactor grows with the increase of steam generators decontaminated during a preventive maintenance

A parametric thermohydraulic study an advanced pressurized light water reactor with a tight fuel rod lattice

International Nuclear Information System (INIS)

Dalle Donne, M.; Hame, W.

1982-12-01

A parametric thermohydraulic study for an Advanced Pressurized Light Water Reactor (APWR) with a tight fuel rod lattice has been performed. The APWR improves the uranium utilisation. The APWR core should be placed in a modern German PWR plant. Within this study about 200 different reactors have been calculated. The tightening of the fuel rod lattice implies a decrease of the net electrical output of the plant, which is greater for the heterogeneous reactor than for the homogeneous reactor. APWR cores mean higher core pressure drops and higher water velocities in the core region. The cores tend to be shorter and the number of fuel rods to be higher than for the PWR. At the higher fuel rod pitch to diameter ratios (p/d) the DNB limitation is more stringent than the limitation on the fuel rod linear rating given by the necessity of reflooding after a reactor accident. The contrary is true for the lower p/d ratios. Subcooled boiling in the highest rated coolant channels occurs for the most of the calculated reactors. (orig.) [de

Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

International Nuclear Information System (INIS)

Marsodi; Mulyanto; Kitamoto, Asashi.

1994-01-01

Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

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.

Installation and testing of the ERANOS computer code for fast reactor calculations

International Nuclear Information System (INIS)

Gren, Milan

2010-12-01

The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

Mathematical foundation of the application of modal analysis to the investigation of space-time reactor behaviour

International Nuclear Information System (INIS)

Obradovic, D.M.

1970-01-01

In recent years investigations in the field of kinetics and dynamics of nuclear reactors have been directed towards overcoming an insufficiently accurate point reactor model. For that purpose different mathematical approaches have been used. This thesis is devoted to modal analysis because, from the practical point of view, it is a very promising and, from the mathematical and physical point of view, a very interesting method. Some fundamental mathematical problems connected with the application of modal analysis to the investigations of the reactor space-time behaviour are still unsolved and accordingly our purpose is to solve some of these problems. The spectral properties of the diffusion and P 1 operators are studied in some detail applying the Krein-Rutman theory of the K-positive operators, the Krasnosel'skii theory of u 0 operators, and the Keldis theory of the operator families. The formal solution to the initial value problem (as an abstract Cauchy problem), associated with the diffusion and P 1 operators is also studied. Modal analysis is identified as a set of methods in the mathematical literature known as the Galerkin methods (or projection methods). Following this idea (using the results of the mathematical investigations of the Galerkin methods) and using our results of the investigations of the properties of the diffusion and P 1 operators, the applicability of modal analysis to the approximate solution of the diffusion and P 1 equations and of the eigenvalue problems associated with the diffusion and P 1 operators is established. As an example of the application of modal analysis the Bubnov and Galerkin method is applied to a multiregion thermal nuclear reactor for the determination of: (i) frequency response, (ii) eigenvalues and eigenvectors of the stationary diffusion operator, (iii) eigenvalues and eigenvectors of the non-stationary diffusion operators. On the basis of the expressions obtained the corresponding computer programmes for radial

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)

Two Stage Anaerobic Reactor Design and Treatment To Produce Biogas From Mixed Liquor of Vegetable Waste

Science.gov (United States)

Budiastuti, H.; Ghozali, M.; Wicaksono, H. K.; Hadiansyah, R.

2018-01-01

Municipal solid waste has become a common challenged problem to be solved for developing countries including Indonesia. Municipal solid waste generating is always bigger than its treatment to reduce affect of environmental pollution. This research tries to contribute to provide an alternative solution to treat municipal solid waste to produce biogas. Vegetable waste was obtained from Gedebage Market, Bandung and starter as a source of anaerobic microorganisms was cow dung obtained from a cow farm in Lembang. A two stage anaerobic reactor was designed and built to treat the vegetable waste in a batch run. The capacity of each reactor is 20 liters but its active volume in each reactor is 15 liters. Reactor 1 (R1) was fed up with mixture of filtered blended vegetable waste and water at ratio of 1:1 whereas Reactor 2 (R2) was filled with filtered mixed liquor of cow dung and water at ratio of 1:1. Both mixtures were left overnight before use. Into R1 it was added EM-4 at concentration of 10%. pH in R1 was maintained at 5 - 6.5 whereas pH in R1 was maintained at 6.5 - 7.5. Temperature of reactors was not maintained to imitate the real environmental temperature. Parameters taken during experiment were pH, temperature, COD, MLVSS, and composition of biogas. The performance of reactor built was shown from COD efficiencies reduction obtained of about 60% both in R1 and R2, pH average in R1 of 4.5 ± 1 and R2 of 7 ± 0.6, average temperature in both reactors of 25 ± 2°C. About 1L gas produced was obtained during the last 6 days of experiment in which CH4 obtained was 8.951 ppm and CO2 of 1.087 ppm. The maximum increase of MLVSS in R1 reached 156% and R2 reached 89%.

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M S.T.; Salakana, L K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1998-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1997-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

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

Comparisons of TRAC-PF-1 calculations with semiscale Mod-3 small-break tests S-SB-P1 and S-SB-P7

International Nuclear Information System (INIS)

Sahota, M.S.

1982-01-01

Semiscale Tests S-SB-P1 and S-SB-P7 conducted in the Semiscale Mod-3 facility at the Idaho National Engineering Laboratory are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1). The results are used to assess TRAC-PF1 predictions of thermal-hydraulic phenomena and the effects of break size and pump operation on system response during slow transients. Tests S-SB-P1 and S-SB-P7 simulated an equivalent pressurized-water-reactor (PWR) 2.5% communicative cold-leg break for early and late pump trips, respectively, with only high-pressure injection (HPI) into the cold legs. The parameters examined include break flow, primary-system pressure response, primary-system mass distribution, and core characteristics

Proceedings of the Seminar on Research Result of Research Reactor Technology Centre 2003

International Nuclear Information System (INIS)

Endiah Puji Hastuti; Setiyanto; Taswanda Taryo; Mohammad Dhandhang Purwadi; Pinem, Surian; Tarigan, Alim; Hasibuan, Djaruddin; Kadarusmanto; Amir Hamzah

2004-05-01

The Proceeding of the Seminar on Research Result of Research Reactor Technology Centre 2003 held by P2TRR has been reported researcher are expected to use the reports as references to research activities in Science and Technology, especially in field of Nuclear Reactor. There are 27 papers which have separated index. (PPIN)

Bioreduction of para-chloronitrobenzene in drinking water using a continuous stirred hydrogen-based hollow fiber membrane biofilm reactor

Energy Technology Data Exchange (ETDEWEB)

Xia Siqing, E-mail: siqingxia@gmail.com [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China); Li Haixiang; Zhang Zhiqiang [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China); Zhang Yanhao [College of Municipal and Environmental Engineering, Shandong Jianzhu University, Jinan 250101 (China); Yang Xin; Jia Renyong; Xie Kang; Xu Xiaotian [State Key Laboratory of Pollution Control and Resource Reuse, College of Environmental Science and Engineering, Tongji University, Shanghai 200092 (China)

2011-08-30

Highlights: {yields} We designed a novel hollow fiber membrane biofilm reactor for p-CNB removal. {yields} Biotransformation pathway of p-CNB in the reactor was investigated in this study. {yields} Nitrate and sulfate competed more strongly for hydrogen than p-CNB. {yields} This reactor achieved high removal efficiency and hydrogen utilization efficiency. - Abstract: para-Chloronitrobenzene (p-CNB) is particularly harmful and persistent in the environment and is one of the priority pollutants. A feasible degradation pathway for p-CNB is bioreduction under anaerobic conditions. Bioreduction of p-CNB using a hydrogen-based hollow fiber membrane biofilm reactor (HFMBfR) was investigated in the present study. The experiment results revealed that p-CNB was firstly reduced to para-chloraniline (p-CAN) as an intermediate and then reduced to aniline that involves nitro reduction and reductive dechlorination with H{sub 2} as the electron donor. The HFMBfR had reduced p-CNB to a major extent with a maximum removal percentage of 99.3% at an influent p-CNB concentration of 2 mg/L and a hydraulic residence time of 4.8 h, which corresponded to a p-CNB flux of 0.058 g/m{sup 2} d. The H{sub 2} availability, p-CNB loading, and the presence of competing electron acceptors affected the p-CNB reduction. Flux analysis indicated that the reduction of p-CNB and p-CAN could consume fewer electrons than that of nitrate and sulfate. The HFMBfR had high average hydrogen utilization efficiencies at different steady states in this experiment, with a maximum efficiency at 98.2%.

Parameter analysis calculation on characteristics of portable FAST reactor

International Nuclear Information System (INIS)

Otsubo, Akira; Kowata, Yasuki

1998-06-01

In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

Nonlinear punctual dynamic applied to simulation of PWR type reactors

International Nuclear Information System (INIS)

Cysne, F.S.

1978-01-01

In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

User's manual for THYDE-P1

International Nuclear Information System (INIS)

Asahi, Yoshiro

1982-04-01

THYDE-P1 is a computer code applicable to LWR (light water reactor) plant dynamics in response to various disturbances. This work is the user's mannual of THYDE-P1 (version SV02L03). The input requirements, steady state adjustment, execution of runs and output specifications are described. (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)

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

Energy-averaged neutron cross sections of fast-reactor structural materials

International Nuclear Information System (INIS)

Smith, A.; McKnight, R.; Smith, D.

1978-02-01

The status of energy-averaged cross sections of fast-reactor structural materials is outlined with emphasis on U.S. data programs in the neutron-energy range 1-10 MeV. Areas of outstanding accomplishment and significant uncertainty are noted with recommendations for future efforts. Attention is primarily given to the main constituents of stainless steel (e.g., Fe, Ni, and Cr) and, secondarily, to alternate structural materials (e.g., V, Ti, Nb, Mo, Zr). Generally, the mass regions of interest are A approximately 50 to 60 and A approximately 90 to 100. Neutron total and elastic-scattering cross sections are discussed with the implication on the non-elastic-cross sections. Cross sections governing discrete-inelastic-neutron-energy transfers are examined in detail. Cross sections for the reactions (n;p), (n;n',p), (n;α), (n;n',α) and (n;2n') are reviewed in the context of fast-reactor performance and/or diagnostics. The primary orientation of the discussion is experimental with some additional attention to the applications of theory, the problems of evaluation and the data sensitivity of representative fast-reactor systems

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

Biohydrogen production from diary processing wastewater by anaerobic biofilm reactors

Energy Technology Data Exchange (ETDEWEB)

Rios-Gonzalez, L.J.; Moreno-Davila, I.M.; Rodriguez-Martinez, J.; Garza-Garcia, Y. [Universidad Autonoma de Coahuila, Saltillo, Coahuila (Mexico)]. E-mail: leopoldo.rios@mail.uadec.mx

2009-09-15

This article describes biological hydrogen production from diary wastewater via anaerobic fermentation using pretreated heat shock (100 degrees Celsius, 30 min.) and acid (pH 3.0, 24 h) treatment procedures to selectively enrich the hydrogen producing mixed consortia prior to inoculation to batch reactors. Bioreactor used for immobilization consortia was operated at mesophilic (room) temperature (20{+-}3 degrees Celsius), under acidophilic conditions (pH 4.0-4.5), HRT (2h), and a natural support for generate hydrogen producing mixed consortia biofilm: Opuntia imbricata. Reactor was initially operated with sorbitol (5g/L) for 60 days of operation. Batch tests were conducted using 20{+-}0.02g of natural support with biofilm. Batch experiments were conducted to investigate the effect of COD (2.9-21.1 g-COD/L), at initial pH of 7.0, 32{+-}1 degrees Celsius. Maximum hydrogen yield was obtained at 21.1 g-COD/L. Experiments of pH effect were conducted using the optimal substrate concentration (21.2 g-COD/L), at pH 4 to 7 and 11.32 (pH diary wastewater) ,and 32{+-}1 degrees Celsius. Experiments results indicate the optimum initial cultivation was pH 4.0, but we can consider also a stable hydrogen production at pH 11.32 (pH diary wastewater), so we can avoid to fit the pH, and use diary wastewater as it left the process of cheese manufacture. The operational pH of 4.0 is 1.5 units below that of previously reported hydrogen producing organisms. The influence of the effect of temperature were conducted using the optimal substrate concentration (21.2 g-COD/L), two pH levels: 4.0 and 11.32, and four different temperatures: 16{+-}3 degrees Celsius (room temperature), 3 C, 45{+-}1 degrees Celsius y 55{+-}1 degrees Celsius.Optimal temperature for hydrogen production from diary wastewater at pH 4.0 was 55{+-}1 degrees Celsius, and for pH 11.32 was 16{+-}3 degrees Celsius.Therefore, the results suggests biofilm reactors in a natural support like Opuntia imbricata have good potential

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.)

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.)

V.S.O.P. (99/05) computer code system

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.

2005-11-01

V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code (∼65000 Fortran statements). (orig.)

V.S.O.P. (99/05) computer code system

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.

2005-11-01

V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)

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)

Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes Savannah River Project L and P Reactors

International Nuclear Information System (INIS)

Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P.

1989-01-01

A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. the potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break

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