Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

Application of non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

International Nuclear Information System (INIS)

Silva, Jose Eduardo Rosa da

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

Modernization of control instrumentation and security of reactor IAN - R1

International Nuclear Information System (INIS)

Gonzalez, J. M.

1993-01-01

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

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

Energy Technology Data Exchange (ETDEWEB)

Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A. [CEA, DEN, DER/SPRC/LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

2012-07-01

With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes

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)

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

Development Plan and R and D Status of China Lead-based Reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS [Institute of Nuclear Energy Safety Technology, Beijing (Switzerland)

2013-07-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Development Plan and R and D Status of China Lead-based Reactor

International Nuclear Information System (INIS)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS

2013-01-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

IPR-R1 TRIGA research reactor decommissioning plan

International Nuclear Information System (INIS)

Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

2008-01-01

The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

Lead-based Fast Reactor Development Plan and R&D Status in China

International Nuclear Information System (INIS)

Wu Yican

2013-01-01

• Lead-based fast reactors have good potential for waste transmutation, fuel breeding and energy production, which has been selected by CAS as the advanced reactor development emphasis with the support of ADS program and MFE program. Sharing of technologies R&D is possible among GIF/ADS/Fusion. • The concepts and test strategy of series China lead-based fast reactors (CLEAR) have been developed. The preliminary engineering design and safety analysis of CLEAR-I are underway. • Technology R&D on CLEAR with series lead alloy loops and accelerator-based neutron generator have been constructed or under construction. • CLEAR series reactor design and construction have big challenges, widely international cooperation on reactor design and technology R&D is welcome

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

R and D directions for the development of CANDU reactors

International Nuclear Information System (INIS)

Torgerson, D.F.

1998-01-01

Full text: AECL is carrying out a comprehensive R and D programme to advance all aspects of CANDU reactor technology. These programs are focusing on three main strategic directions: improved economics, enhanced safety, and fuel cycle flexibility. R and D areas include fuel cycle development, heavy water technology, fuel channel development, safety technology, control and instrumentation, reactor chemistry, systems and components, and health and environment. In each case, the R and D programs have short, medium, and long-term goals to achieve the overall strategic directions. Most of the programs seek to further develop and exploit some of the unique characteristics of pressurized heavy water reactors. Examples of this include high neutron economy and on-power fueling which allow several different fuel cycles, the presence of large water heat sinks for enhanced safety, and modular components that can be easily replaced for plant life extension. This presentation reviews AECL's product development directions and the R and D programs that have been begun for their development

Characteristics of biohydrogen production by ethanoligenens R{sub 3} isolated from continuous stirred tank reactor

Energy Technology Data Exchange (ETDEWEB)

Jiao, A.Y.; Liu, K. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering; Liu, B. [Northeast Forestry Univ., Harbin (China). School of Material Science and Engineering; Xu, J.L. [Shanghai Univ. of Engineering Science (China). College of Chemistry and Chemical Engineering

2010-07-01

This study investigated the fermentative hydrogen production characteristics of ethanoligenens R{sub 3} isolated from anaerobic sludge in a continuous stirred tank reactor. The effects of the initial pH value, the proportion of carbon and nitrogen sources, and the effects of fermentation temperature were investigated in a series of batch experiments. Substrates for the hydrogen production of glucose and peptone were used as carbon and nitrogen sources. Results of the experiments showed that a maximum hydrogen production yield of 834 mlH{sub 2}/L culture was obtained with a fermentation temperature of 35 degrees C and an initial pH value of 5.5. The maximum average hydrogen production rate of 10.87 mmolH{sub 2}/g cell dry weight per hour was obtained at a carbon-nitrogen source ratio of 3.3. The degradation efficiency of the glucose used as a carbon source ranged from 91.5 to 95.43 per cent during the conversion of glucose to hydrogen by the bacteria.

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

Decommissioning and decontrolling the R1-reactor

International Nuclear Information System (INIS)

Bergman, C.; Holmberg, B.T.

1985-01-01

Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

The molten salt reactor: R and D status and perspectives in Europe

International Nuclear Information System (INIS)

Renault, Claude; Delpech, Sylvie; Merle-Lucotte, Elsa; Konings, Rudy; Hron, Miloslav; Ignatiev, Victor

2010-01-01

The paper concentrates on molten salt fast reactor (MSFR) concepts which are receiving most attention in the EU context. It shows the main R and D achievements and some remaining issues to be addressed in such essential areas as (a) reactor conceptual design, (b) molten salt properties, (c) fuel salt clean-up scheme and (d) high temperature materials. The status and perspectives of molten salt reactor R and D efforts in Europe are then discussed

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-15

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

An efficient conversion of (3R,3'R,6'R)-lutein to (3R,3'S,6'R)-lutein (3'-epilutein) and (3R,3'R)-zeaxanthin.

Science.gov (United States)

Khachik, Frederick

2003-01-01

Two dietary carotenoids, (3R,3'R,6'R)-lutein (1) and (3R,3'R)-zeaxanthin (2), and their metabolite (3R,3'S,6'R)-lutein (3'-epilutein) (3) accumulate in human serum, milk, and ocular tissues. There is increasing evidence that compounds 1 and 2 play an important role in the prevention of age-related macular degeneration. Therefore, the availability of these carotenoids for metabolic studies and clinical trials is essential. Compound 1 is isolated from extracts of marigold flowers (Tagete erecta) and is commercially available, whereas 2 is only accessible by a lengthy total synthesis, and a viable method for synthesis of 3 has not yet been developed. This report describes an efficient conversion of technical grade 1 to 2 via 3. Acid-catalyzed epimerization of 1 yields an equimolar mixture of diastereomers 1 and 3. The mixture was separated by enzyme-mediated acylation with lipase AK from Pseudomonas fluorescens that preferentially esterified 3 and after alkaline hydrolysis yielded this carotenoid in 90% diastereomeric excess (de). Compound 3 was also separated from 1 in 56-88% de by solvent extraction and low-temperature crystallization, Soxhlet extraction, or supercritical fluid extraction. Base-catalyzed isomerization of 3 gave 2 in excellent yield, providing a convenient alternative to the total synthesis of this important dietary carotenoid.

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

International Nuclear Information System (INIS)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A.

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U 3 O 8 and U 3 Si 2 dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U 3 O 8 -Al and five containing U 3 Si 2 -Al), with densities of 3.2 gU/cm 3 and 4.8 gU/cm 3 respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

R-matrix parameters in reactor applications

International Nuclear Information System (INIS)

Hwang, R.N.

1992-01-01

The key role of the resonance phenomena in reactor applications manifests through the self-shielding effect. The basic issue involves the application of the microscopic cross sections in the macroscopic reactor lattices consisting of many nuclides that exhibit resonance behavior. To preserve the fidelity of such a effect requires the accurate calculations of the cross sections and the neutron flux in great detail. This clearly not possible without viable resonance data. Recently released ENDF/B VI resonance data in the resolved range especially reflect the dramatic improvement in two important areas; namely, the significant extension of the resolved resonance ranges accompanied by the availability of the R-matrix parameters of the Reich-Moore type. Aside from the obvious increase in computing time required for the significantly greater number of resonances, the main concern is the compatibility of the Riech-Moore representation to the existing reactor processing codes which, until now, are based on the traditional cross section formalisms. This purpose of this paper is to summarize our recent efforts to facilitate implementation of the proposed methods into the production codes at ANL

Environmental issues associated with R Reactor renovation and startup

International Nuclear Information System (INIS)

Marter, W.L.

1982-01-01

This memorandum identifies the more significant environmental issues that would be associated with renovation and startup of R Reactor to meet future demands for nuclear weapon materials. Some key environmental issues identified are: potential occupational radiation exposures associated with repairing the leaks in the reactor nozzles and thermal shield tanks and with renovating the disassembly basin facilities; the impacts of constructing additional cooling water pumping capacity; the effect of increased pumping on impingement and entrainment at the pump houses; thermal effects on the R Canal-precooler pond-Par Pond ecosystem from increased discharge of cooling water; and effects of increased water flow in Lower Three Runs Creek on biota and on cesium-137 remobilization

Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2002-07-01

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)

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Partial synthesis of (3R,6'R)-alpha-cryptoxanthin and (3R)-beta-cryptoxanthin from (3R,3'R,6'R)-lutein.

Science.gov (United States)

Khachik, Frederick; Chang, An-Ni; Gana, Audry; Mazzola, Eugene

2007-02-01

(3R,3'R,6'R)-Lutein (1), (3R,3'R)-zeaxanthin (2), (3R,6'R)-alpha-cryptoxanthin (3), and (3R)-beta-cryptoxanthin (4) are among dietary hydroxycarotenoids that have been identified in human serum, milk, and ocular tissues. While 1 containing 6% of 2 is commercially available, industrial production of optically active 3 and 4 has not yet been accomplished. Several processes have been developed that transform 1 into 3, 4, and minor quantities of (3R,5'RS,6'R)-3',4'-didehydro-5',6'-dihydro-beta,beta-caroten-3-ol (5) (a regioisomer of 3). In one process, lutein (1) was cleanly deoxygenated to 3 in the presence of trifluoroacetic acid (TFA) and Me3N.BH3 in CH2Cl2 at ambient temperature in nearly 90% yield. Reaction of lutein (1) with a Lewis acid (AlCl3, ZnBr2, ZnI2) and a hydride donor (Me3N.BH3, Na[BH3(OCOCF3)], NaCNBH3) in solvents such as CH2Cl2, THF, and TBME produced similar results. In a two-step process, high-temperature acid-catalyzed dehydration of 1 (propanol/water/acid, 90 degrees C) gave a mixture of anhydroluteins 6, 7, and 8 in 86% yield. In the second step, these dehydration products underwent ionic hydrogenation with TFA/Me3N.BH3 in CH2Cl2 to afford a mixture of 3 and 4 in nearly 80% yield that contained only 1% of 5.

Neutron radiography in the IEA-R1 reactor

International Nuclear Information System (INIS)

Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

1988-08-01

Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

Studsvik's R2 reactor - Review of activities

Energy Technology Data Exchange (ETDEWEB)

Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

1993-07-01

A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

Management of European fast reactor R and D

International Nuclear Information System (INIS)

Judd, A.M.; Sheriff, N.

1993-01-01

Since 1984 government-funded fast reactor R and D in France, Germany and the UK has been run as a collaborative activity, and since 1988 as a unified programme in support of the design and construction of the advanced European Fast Reactor. This paper describes the international management structure which has been set up, and the means used to control the work. It is written from the point of view of those engaged in the project, and makes no attempt at a formal analysis of the structure. The main difficulty is that control of funding remains with the three governments. The R and D programme has to be managed so that it meets the needs of each government separately as well as the designers' requirements. To start with the management structure was excessively bureaucratic, but it has become more flexible and efficient. This has happened as the initial nationalistic suspicions have broken down, and the staff engaged in the work have learnt more about each others' ways of working so that an atmosphere of trust and inter-dependence has grown up. (This paper was written before the changes in UK policy on fast reactor development were announced in November 1992). (Author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Fast reactors: R and D targets and outlook for their introduction

International Nuclear Information System (INIS)

Poplavsky, V.; Barre, B.; Aizawa, K.

1997-01-01

In this paper the current status of fast reactors development is briefly outlined, including experimental, demonstration, and commercial installations. Data on the experience gained in development and operation of NPPs with reactors of this type are presented. The issues are discussed in connection with possibilities of fast reactor development in the nuclear power structure for the near (up to 2010-2020) and distant future. In the final part of the paper, an analysis is given of possible ways for R and D development in the field of NPPs with fast neutron reactors. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter

2015-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

R and D on fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

2012-01-01

Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

New digital control system for the operation of the Colombian research reactor IAN-R1

International Nuclear Information System (INIS)

Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

2015-09-01

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

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

Comparison of ASTEC 1.3 and ASTEC 1.3 R2 calculations in case of SBO for VVER-1000 reactor

International Nuclear Information System (INIS)

Atanasova, B.; Stefanova, A.; Grudev, P.

2009-01-01

The report presents the results from severe accident analyses performed with the both versions of ASTEC v1.3 and ASTEC v1.3R2 computer code for a VVER 1000 type of reactor. The purpose of this analysis is to assess the progress of ASTEC code modeling of main phenomena arising during hypothetical severe accidents. The final target of these analyses is to estimate the behaviour of the ASTEC code, its capability for simulation of severe accidents, including safety systems and Severe Accident Management (SAM) procedures. The analyses have been performed assuming a station blackout with simultaneous loss of HPIS, LPIS (ECCSs), EFWS and spray system due to failure of DGs. Hydro accumulators are not available. In the calculation it is assumed opening and stuck-open of PRZ relief valves. It has been organized the Fission Products path through the SEMPELL valve. It should be said that this investigation was limited to the 'in-vessel' phase of the sequence; therefore the effect of sprays on containment atmosphere has not been studied. (authors)

Modifications done in the IPR-R1 reactor and their auxiliary systems

International Nuclear Information System (INIS)

Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

1986-01-01

The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

Experience and research with the IEA-R1 Brazilian reactor

International Nuclear Information System (INIS)

Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

1982-06-01

The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

Design and properties of marine reactors and associated R and D

Energy Technology Data Exchange (ETDEWEB)

Gagarinski, A; Ignatiev, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

Design and properties of marine reactors and associated R and D

International Nuclear Information System (INIS)

Gagarinski, A.; Ignatiev, V.; Devell, L.

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Radioactive inventory in structural materials of ET-R R-1 reactor and its implication on decommissioning.

Energy Technology Data Exchange (ETDEWEB)

Elkady, A; Amin, E [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

A plan for decommissioning of ET-R R-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences decommissioning. Conservative calculations have been made to evaluate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are present in significant quantities in the reactor structural materials are aluminium, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from Co-60 and Fe-55 which are present in aluminium as trace elements in larger quantities in other construction materials. 2 figs., 4 tabs.

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Energy Technology Data Exchange (ETDEWEB)

Wiersma, B.

2010-05-24

operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

Energy Technology Data Exchange (ETDEWEB)

Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

2007-08-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

Current activities at the FiR 1 TRIGA reactor

International Nuclear Information System (INIS)

Salmenhaara, Seppo

2002-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

The Canadian R and D program targeted at CANDU reactors

International Nuclear Information System (INIS)

Moeck, E.O.

1988-01-01

CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

International Nuclear Information System (INIS)

Paiva, Rosemeire P.; Salvetti, Tereza C.

2005-01-01

Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

International Nuclear Information System (INIS)

Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

2006-01-01

This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

Nuclear material control at IEA-R1 nuclear research reactor

International Nuclear Information System (INIS)

1988-01-01

The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

Modification of the IAN-R1 reactor

International Nuclear Information System (INIS)

Jaime, J.; Ahumada, S.; Spin, R.A.

1990-01-01

The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

ORNL R and D on advanced small and medium power reactors: selected topics

International Nuclear Information System (INIS)

White, J.D.; Trauger, D.B.

1989-01-01

From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, an assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable R and D would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current R and D efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described

Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

International Nuclear Information System (INIS)

Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

2010-01-01

In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

Strengthening the R and D on fast reactor technology, and promoting its industrialization

International Nuclear Information System (INIS)

Wan Gang

2008-01-01

Based on the strategic thoughts of energy development in China expounded by Jiang Zemin in the article entitled 'Reflections on Energy Issues in China', the author points out in this paper that R and Ds on fast reactor technology shall be carried out timely in China by taking full advantage of international scientific resources, and overall planning in this regard shall be made as well. The point of view of strengthening fast reactor technology R and D and promoting its industrialization is also put forward in the paper. (authors)

Commissioning of the new heat exchanger for the research nuclear reactor IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo Jose Alvim de; Cassiano, Douglas Alves; Umbehaun, Pedro Ernesto; Carvalho, Marcos Rodrigues de; Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ajcastro@ipen.br; docass@gmail.com.br; umbehaun@ipen.br; carvalho@ipen.br; frajndli@ipen.br

2008-07-01

The Research Reactor IEA-R1 placed at IPEN/CNEN-SP is of the swimming pool type, light water moderated and with graphite reflectors, and was build and designed by Babcock and Wilcox Co. Start up operation was in September the 16{sup th}, 1957, being the first criticality for South Hemisphere. Although designed to operate at 5 MW, the IEA-R1 was operated until 2001 with 2 MW and was suitable for use in basic and applied research as well as the production of medical radioisotopes, industry and natural sciences applications. Due to a recent demand increase on radioisotopes in Brazil for medical diagnoses and therapies applications, IPEN /CNEN updated the IEA-R1 power to 5 MW and to work at continuous operation regime. Studies on the Ageing Management for the Research Reactor IEA-R1 were conducted according to IAEA procedures. As result of these studies critical components within the Ageing Management Program were identified. Also were made recommendations on the implementation of test scheduling and standardization procedures to organize data and documents. One of the main results was the need of monitoring the two heat exchangers, the two primary circuit pumps and the data acquisition system. During monitoring procedures, issues were observed on the IEA-R1 operation at 5 MW mainly due to the ageing of the Babcox and Wilcox TCA heat exchanger, and excessive vibrations at high flow rates on CBC's TCB heat exchanger. So, from 2005 on, it was decided to work with 3,5 MW and provide a new IESA heat exchanger with 5 MW capacity, to substitute the TCA heat exchanger. This work presents results on the commissioning of the new heat exchanger and compares against the values calculated in the IESA project. The results show that the IEA-R1 Reactor can be operated more safety and continuously at 5 MW with the new IESA heat exchanger. (author)

Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias

2005-01-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

International Nuclear Information System (INIS)

Auterinen, Iiro; Salmenhaara, Seppo

2008-01-01

FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

2009-01-01

The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

Safety approach and R and D program for future french sodium-cooled fast reactors

International Nuclear Information System (INIS)

Beils, Stephane; Carluec, Bernard; Devictor, Nicolas; Fiorini, Gian Luigi; Sauvage, Jean Francois

2011-01-01

This paper presents briefly the safety approach as well as the R and D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R and D. B) Strategy and roadmap in support of the R and D for future SFRs. This section describes the R and D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype. (author)

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

Trends on R and D of the innovative nuclear reactors in Japan

International Nuclear Information System (INIS)

Kinoshita, Izumi

2002-01-01

In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

Process modeling of a reversible solid oxide cell (r-SOC) energy storage system utilizing commercially available SOC reactor

International Nuclear Information System (INIS)

Mottaghizadeh, Pegah; Santhanam, Srikanth; Heddrich, Marc P.; Friedrich, K. Andreas; Rinaldi, Fabio

2017-01-01

Highlights: • An electric energy storage system was developed based on a commercially available SOC reactor. • Heat generated in SOFC mode of r-SOC is utilized in SOEC operation of r-SOC using latent heat storage. • A round trip efficiency of 54.3% was reached for the reference system at atmospheric pressure. • An improved process system design achieved a round-trip efficiency of 60.4% at 25 bar. - Abstract: The increase of intermittent renewable energy contribution in power grids has urged us to seek means for temporal decoupling of electricity production and consumption. A reversible solid oxide cell (r-SOC) enables storage of surplus electricity through electrochemical reactions when it is in electrolysis mode. The reserved energy in form of chemical compounds is then converted to electricity when the cell operates as a fuel cell. A process system model was implemented using Aspen Plus® V8.8 based on a commercially available r-SOC reactor experimentally characterized at DLR. In this study a complete self-sustaining system configuration is designed by optimal thermal integration and balance of plant. Under reference conditions a round trip efficiency of 54.3% was achieved. Generated heat in fuel cell mode is exploited by latent heat storage tanks to enable endothermic operation of reactor in its electrolysis mode. In total, out of 100 units of thermal energy stored in heat storage tanks during fuel cell mode, 90% was utilized to offset heat demand of system in its electrolysis mode. Parametric analysis revealed the significance of heat storage tanks in thermal management even when reactor entered its exothermic mode of electrolysis. An improved process system design demonstrates a system round-trip efficiency of 60.4% at 25 bar.

Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

International Nuclear Information System (INIS)

Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

2013-06-01

(Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

Development of a training simulator to operators of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carvalho, Ricardo Pinto de

2006-01-01

This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

Auxiliary control system of the safety parameters for IPR-R1 reactor

International Nuclear Information System (INIS)

Coura, J.G.

1986-01-01

This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Spent fuel management - two alternatives at the FiR 1 reactor

International Nuclear Information System (INIS)

Salmenhaara, S.E.J.

2001-01-01

The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

Core calculations for the upgrading of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

1998-01-01

The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

International Nuclear Information System (INIS)

D'Utra Bitelli, U.

1993-01-01

This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

New burnup calculation of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

2015-01-01

The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

2013-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

A total Ammonium Reactor (NHxR) for In Situ Mobile Measurements: A Critical Tool to Understand Aerosol Formation, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — We will develop, demonstrate, and optimize a front-end ammonium reactor (NHxR) for the fast, precise, and accurate measurement of gas-phase ammonia (NH3) and...

Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2016-01-01

The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

Spent fuel management plans for the FiR 1 Reactor

International Nuclear Information System (INIS)

Salmenhaara, S. E. J.

2002-01-01

The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

Safety-related Innovative Nuclear Reactor Technology Elements R and D (SINTER) Network and Global HTGR R and D Network (GHTRN). Strategic benefits of international networking

International Nuclear Information System (INIS)

Von Lensa, W.

1998-01-01

Action on 'Safety-related Innovative Nuclear Reactor Technology Elements - R and D - (SINTER) Network' both aim at the identification of priority items for sustainable innovations of nuclear technologies and work-shared European collaboration structures. Such an approach can also be realised for future R and D on HTGR-related R and D under the umbrella of the IAEA as already proposed by the 'International Working Group on Gas-Cooled Reactors (IWGGCR)' in 1996 and illustrated in this paper for the construction of a 'Global HTGR R and D Network (GHTRN)'. 3 refs

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

International Nuclear Information System (INIS)

Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

1976-04-01

The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

Measurements and calculations of reactivity for the IEA-R1 reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

1988-01-01

This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

International Nuclear Information System (INIS)

Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

2013-01-01

The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

International Nuclear Information System (INIS)

Duque O, J.; Velez A, L.H.

1987-01-01

Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

2009-01-01

The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.

1988-01-01

Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

Calculation of the main neutron parameters of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Ojima, Mario Katsuhiko

1977-01-01

The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

2014-08-15

The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

Neutronic, thermal-hydraulics and accident analysis calculations for an irradiation device to be used in the qualification process of dispersion fuels in the IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br

2009-07-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 235}U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor, now in the conception phase. For the neutronic calculation, the computer code CITATION was utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)

Status of the R and D activities on fast reactors and ADS in Brazil

International Nuclear Information System (INIS)

Maiorino, Jose Rubens

2001-01-01

Research and Development in Nuclear Science and Technology is conducted by Research Institutes of the Brazilian Nuclear Energy Commission. In Fast Reactor, R and D activities started in the sixties, and in 1972 a small Na loop (100 kW) was constructed. At the same time, during the seventies at IPEN, research in cooperation with GA for Gas Cooled Fast Breeder Reactor was conducted. The motivation of such research was Thorium Fuel Cycle. As a result of this research a Helium Loop was constructed and a Split Table Critical Assembly (ZPR) was designed. During the eighties, an agreement with ANSALDO-NIRA resulted in an acquisition of a Sodium Loop for Thermohydraulics studies, however it never had been assembled. At the same time, a concept of a Binary Breeder Reactor using two cycles, Th and U, was developed. During the nineties, a National Program to conduct R and D (pyroprocess; U-Zr Metallic Fuel; HT-9; Electromagnetic Pump; and a conceptual design of a Experimental Reactor (60/20 MWth/MWe)) was proposed, however it was closed at the end of the decade. Now, only academic research is being conducted, and it is summarized in this report. Basically, they are: an integral lead fast reactor concept for developing countries, and an alternative concept for a fast energy amplifier accelerator driven system. The first is an combination of best characteristics of the American Integral Fast Reactor and the Russian Lead Cooled Reactor. The second is a conceptual design of ADS helium cooled imbedded in a solid lead subcritical array of fuel, using more than one point of spallation trying to reduce the requirement for energy and current of the accelerator

Optimization of neutronic characteristics of U3Si2 low enrichment fuel elements for a new design of IEA-R1 reactor core

International Nuclear Information System (INIS)

Mai, L.A.; Maiorino, J.R.; Gouvea, E.A.

1989-01-01

This work shows a study of neutronic optimization of U 3 Si 2 -Al low enrichment fuel element. This study has a goal to propose a optimized Core to be used in the research reactor IEA-R1. The external dimensions of the fuel element were maintained as constraints and the loss of reactivity along fuel life-time was defined as 'objective function', and it has been minimized by varying the fuel element dimensions. Cell calculations were made with HAMMER-TECH /3/ Code, for burnups up to 50% of U-235 initial mass. The Computer values of the objective function for several combinations of fuel element dimensions were fitted by a surface using the SAS system /9/, and it has been minimized by a Harwell subroutine /10/. (author) [pt

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

2002-01-01

Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Modernization of the CDTN IPR-R1 TRIGA reactor instrumentation and control

International Nuclear Information System (INIS)

Mesquita, A.Z.; Costa, A.C.L.; Souza, R.M.G.P.

2009-01-01

The control system of the IPR-R1 was changed in 1995. Although since the year's 80 was generalized the use of microprocessor technology and video monitors for visual interface, in the IPR-R1 control room it was used analogical system by relay-based logic, and were maintained the mechanical strip chart recorders (ink-pen drive) to measure, monitor and store the operational parameters. It was maintained the measure and the control of, practically, the same variables of the original system, although the reactor power already have been upgraded to 100 kW and began the studies to increase it to 250 kW, which is the current core configuration. For 250 kW operations the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art instrumentation and control system using microprocessor technology is proposed to replace the present analogical systems. The new system can eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards as NRC and IEE specifications. This paper describes the research project in process in CDTN that has as objective the modernization of the IPR-R1 TRIGA reactor instrumentation and control of the operational variables. The project also will improve the accomplishment of neutronic and thermal-hydraulic experiments, foreseen in the CDTN research program. (author)

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Aspects of the Iea-R1 research reactor seismic evaluation

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1996-01-01

Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

R and D relative to the serious accidents in the PWR type reactors: assessment and perspectives; R and D relative aux accidents graves dans les reacteurs a eau pressurisee: bilan et perspectives

Energy Technology Data Exchange (ETDEWEB)

Bentaib, A.; Bonneville, H.; Caroli, H.; Chaumont, B.; Clement, B.; Cranga, M.; Koundy, V.; Laurent, B.; Micaelli, J.C.; Meignen, R.; Pichereau, F.; Plassart, D.; Van-Dorsselaere, P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Clamart (France); Ducros, G.; Journeau, Ch.; Magallon, D. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Durin, M.; Studer, E. [CEA Saclay 91 - Gif sur Yvette (France); Seiler, J.M. [CEA Grenoble, 38 (France); Ranval, W. [Electricite de France (EDF), 75 - Paris (France)

2006-07-01

This document presents the current state of the research relative to the grave accidents realized in France and abroad. It aims at giving the most exhaustive possible and objective vision of this original field of research. He allows to contribute to the identification and to the hierarchical organization of the needs of R and D, this hierarchical organization in front of, naturally, to be completed by a strong lighting on needs in terms of safety analyses associated with the different risks and the physical phenomena, in particular with the support of probability evaluations of safety level 2, whose the level of sharpness must be sufficient not to hide, by construction, physical phenomena of which the limited knowledge leads to important uncertainties. Let us note that neither the safety analyses, nor the E.P.S. 2 are presented in this document. This report presents the physical phenomena which can arise during a grave accident, in the reactor vessel and in the reactor containment, their chain and the means allowing to ease the effects. The corresponding scenarios are presented to the chapter 2. The chapter 3 is dedicated to the progress of the accident in the reactor vessel; the degradation of the core in reactor vessel (3.1), the behavior of the corium in bottom of reactor vessel (3.2) the break of the reactor vessel (3.3) and the fusion in pressure (3.4) are thus handled there. The chapter 4 concerns the phenomena which can lead to a premature failure of the containment, namely the direct heating of gases of the containment (4.1), the hydrogen risk (4.2) and the vapor explosion (4.3). The phenomenon which can lead to a delayed failure from the containment, namely the interaction corium-concrete, is approached on the chapter 5. The chapter 6 is dedicated to the problems connected to the keeping back and to the corium cooling in reactor vessel and out of reactor vessel, namely the keeping back in reactor vessel by re-flooding of the primary circuit or by re

2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

International Nuclear Information System (INIS)

Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

1969-01-01

1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

Science.gov (United States)

2012-05-03

... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

International Nuclear Information System (INIS)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

2011-01-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2011-07-01

The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

Science.gov (United States)

2012-11-15

... Research Institute TRIGA Reactor: Facility Operating License No. R-84 AGENCY: Nuclear Regulatory Commission... considering an application for the renewal of Facility Operating License No. R-84 (Application), which... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

Energy Technology Data Exchange (ETDEWEB)

Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

2014-08-15

IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

Study and project of the new rack with boron for storage of fuel elements burned in the IEA-R1 research reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Silva, Davilson Gomes da

2017-01-01

The IEA-R1 research reactor works 40h weekly with 4.5 Mw power. The storage rack for spent fuel elements has less than half of its initial capacity. Under these conditions (current conditions of reactor operation 32h weekly will have 3 spend fuel by year, then, approximately 3 utilization rate Positions/year). Thus, we will have only about six years of capacity for storage. Whereas the desired service life of the IEA-R1 is at least another 20 years, it will be necessary to increase the storage capacity of spent fuel. Hence, it is necessary to double the wet storage capacity (storage in the IEA-R1 reactor's pool). After reviewing the literature about materials available for use in the construction of the new storage rack with absorber of neutrons, the BoralcanTM (manufactured by 3TMhis) was chosen due to its properties. This work presents studies: (a) for the construction of new storages racks with double of the current capacity using the same place of current storages racks and (b) criticality analysis using the MCNP-5 code. Two American Nuclear Data Library were used: ENDF / B-VI and ENDF / B-VII, and the results obtained for each data bases were compared. These analyzes confirm the possibility of doubling the storage capacity of fuel elements burned in the same place occupied by the current storage rack attending to the IEA-R1 reactor needs and attending the safety requirements according to the National Nuclear Energy Commission - CNEN and the International Atomic Energy Agency (IAEA). To calculate the k eff were considered new fuel elements (maximum possible reactivity) used in full charge of the storage rack. With the results obtained in the simulation we can conclude that doubling the amount of racks for spent fuel elements are complied with safety limits established in the IAEA standards and CNEN of criticality (keff < 0.95). (author)

Study and project of the new rack with boron for storage of fuel elements burned in the IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Silva, Davilson Gomes da, E-mail: acirodri@ipen.br, E-mail: tmfilho@usp.br, E-mail: dgsilva@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

The IEA-R1 research reactor works 40h weekly with 4.5 Mw power. The storage rack for spent fuel elements has less than half of its initial capacity. Under these conditions (current conditions of reactor operation 32h weekly will have 3 spend fuel by year, then, approximately 3 utilization rate Positions/year). Thus, we will have only about six years of capacity for storage. Whereas the desired service life of the IEA-R1 is at least another 20 years, it will be necessary to increase the storage capacity of spent fuel. Hence, it is necessary to double the wet storage capacity (storage in the IEA-R1 reactor's pool). After reviewing the literature about materials available for use in the construction of the new storage rack with absorber of neutrons, the BoralcanTM (manufactured by 3TMhis) was chosen due to its properties. This work presents studies: (a) for the construction of new storages racks with double of the current capacity using the same place of current storages racks and (b) criticality analysis using the MCNP-5 code. Two American Nuclear Data Library were used: ENDF / B-VI and ENDF / B-VII, and the results obtained for each data bases were compared. These analyzes confirm the possibility of doubling the storage capacity of fuel elements burned in the same place occupied by the current storage rack attending to the IEA-R1 reactor needs and attending the safety requirements according to the National Nuclear Energy Commission - CNEN and the International Atomic Energy Agency (IAEA). To calculate the k{sub eff} were considered new fuel elements (maximum possible reactivity) used in full charge of the storage rack. With the results obtained in the simulation we can conclude that doubling the amount of racks for spent fuel elements are complied with safety limits established in the IAEA standards and CNEN of criticality (keff < 0.95). (author)

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

2015-07-01

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

Microstructure in Zircaloy Creep Tested in the R2 Reactor

International Nuclear Information System (INIS)

Pettersson, Kjell

2004-12-01

Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

R and D status of an integral type small reactor MRX in JAERI

International Nuclear Information System (INIS)

Hoshi, Tsutao; Ochiai, Masaaki; Iida, Hiromasa; Yamaji, Akio; Shimazaki, Junya

1995-01-01

JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety study of the thermal hydraulic phenomena as well as the feasibility study of the applicability to merchant ships. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the results of feasibility study. (author)

Applications of neutron activation analysis technique in the IPR-R1 research reactor

International Nuclear Information System (INIS)

Sabino, C.V.S.; Mansur, N.

1986-01-01

A review is made of the neutron activation analysis technique used in the IPR-R1 reactor of the Centro de Desenvolvimento da Tecnologia Nuclear - NUCLEBRAS. Some characteristics of the method are described, types of samples and elements analyzed are also mentioned. (Author) [pt

GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

International Nuclear Information System (INIS)

Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

2006-01-01

A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

Present status and future perspective of R and D on lead heavy metal-cooled fast reactors

International Nuclear Information System (INIS)

Takahashi, Minoru

2007-01-01

Since a lead heavy metal (lead-bismuth eutectic) is chemically inert and has higher boiling point compared to a sodium, a lead heavy metal-cooled fast reactor can be inherently safe and has good nuclear characteristics and is so suitable to a medium-small size of the reactor. R and D on corrosion of a lead heavy metal has been carried out in the world and this issue might be solved to choose specific corrosion resistant alloys for structural materials and fuel cans of a lead heavy metal-cooled reactor. This article reviews present status and future perspective on lead heavy metal-cooled fast reactors. (T. Tanaka)

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

Verification of the code DYN3D/R with the help of international benchmarks

International Nuclear Information System (INIS)

Grundmann, U.; Rohde, U.

1997-10-01

Different benchmarks for reactors with quadratic fuel assemblies were calculated with the code DYN3D/R. In this report comparisons with the results of the reference solutions are carried out. The results of DYN3D/R and the reference calculation for the eigenvalue k eff and the power distribution are shown for the steady-state 3-dimensional IAEA-Benchmark. The results of NEACRP-Benchmarks on control rod ejections in a standard PWR were compared with the reference solutions published by the NEA Data Bank. For assessing the accuracy of DYN3D/R results in comparison to other codes the deviations to the reference solutions are considered. Detailed comparisons with the published reference solutions of the NEA-NSC Benchmarks on uncontrolled withdrawal of control rods are made. The influence of the axial nodalization is also investigated. All in all, a good agreement of the DYN3D/R results with the reference solutions can be seen for the considered benchmark problems. (orig.) [de

Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Zacarias

2005-07-01

The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

Méthodologie de l'extrapolation des réacteurs chimiques Methodology for Scaling Up Chemical Reactors

Directory of Open Access Journals (Sweden)

Trambouze P.

2006-11-01

Full Text Available Après un exposé général relatif à la méthodologie du développement des procédés, applicable à l'extrapolation des réacteurs, est présenté un rapide examen critique des deux principales techniques mises en oeuvre, à savoir : - la théorie de la similitude ; - l'élaboration de modèles mathématiques. Deux exemples pratiques, relatifs aux réacteurs homogènes et aux réacteurs catalytiques à lit fixe et deux phases fluides, sont ensuite examinés à la lumière des considérations générales précédentes. After giving a general description of process-development methodology applicable to scaling up reactors, this article makes a quick critical examination of the two main techniques involved, i. e. : (a the theory of similarity, and (b the compiling of mathematical models. Two practical examples relating to homogeneous reactors and trickle-bed catalytic reactors are then examined in the light of the preceding general considerations.

Development of an artificial neural network for nuclear power monitoring and fault detection in the IEA-R1 research reactor at IPEN

International Nuclear Information System (INIS)

Bueno, Elaine Inacio; Ting, Daniel Kao Sun; Goncalves, Iraci M.P.

2005-01-01

The purpose of this paper is to develop a system to monitor the nuclear power of a reactor using Artificial Neural Networks. The database used in this work was developed using a theoretical model of IEA-R1 Research Reactor. The IEA-R1 is a pool type reactor of 5 MW, cooled and moderated by light water, and uses graphite and beryllium as reflector. To monitor the nuclear power the following variables were chosen: T3 . temperature above the reactor core, T4 . outlet core temperature, FE01 . primary loop flow rate and the nuclear power. The inputs are T3, T4 and FE01 and the output is the nuclear power. It was used several networks using the backpropagation algorithm. The conclusion is that the multiplayer perceptrons networks (MLPs), training by the backpropagation algorithm, can be used to solve this problem. The results obtained with the MLPs networks are satisfactory and the mean square error was in the order of 10 -4 during the network training and in the order of 10 -2 during the network testing. We intend to monitor the other variables of this model using the same methodology, and after this we will use the real database from the system to compare the results obtained with the model. The monitoring of the reactor variables is part of the development of a fault detection and isolation system which is underway and which is, by its turn, part of a comprehensive ageing management program. (author)

Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

International Nuclear Information System (INIS)

Moreira, Joao Manoel Losada; Frajndlich, Roberto

1995-01-01

This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

The Effect of Organic Loading Rate on Milk WastewaterTreatment Using Sequencing Batch Reactor (SBR

Directory of Open Access Journals (Sweden)

Hooman Hajiabadi

2009-09-01

Full Text Available In this study, four aerobic sequencing batch reactors (SBRs were operated under the same conditions for the treatment of milk wastewater at different organic loading rates (OLRs. Cylindrical Plexiglas reactors were run for 56 days (including 21 days of acclimatization and 35 days of data gathering. Effective volume, influent wastewater flowrate, and sludge retention time (SRT of reactors were 5.5 L, 3.5 L/d, and 10 d, respectively. The average COD removal efficiency for the reactors R1, R2, R3, and R4 with influent OLRave values of 633, 929, 1915, and 3261 gCOD/m3d were 95, 96, 95, and 82 percent, respectively. The average effluent suspended solid (SS for all reactors was lower than 44 mg/L. Also, except for R4 with an average effluent turbidity of 270 NTU, other reactors met the Iranian wastewater emission standard (50 NTU. In addition, the average sludge volume index of reactors R1 to R3 was found to be lower than 67 mL/g. According to the results, the overall variation of COD removal efficiency versus influent OLR shows a decreasing rate with a correlation factor of 0.8 (R2.

25th birthday of the first criticality of IPR-R1 nuclear research reactor

International Nuclear Information System (INIS)

Tofani, P.C.; Stasiulevicius, R.; Roedel, G.

1988-01-01

The historical evolution of IPR-R1 research reactor of Instituto de Pesquisas Radioativas-Nuclebras, since the data of its first criticality, is presented. The modifications and the main activities carried out, are presented. (M.C.K.) [pt

Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

International Nuclear Information System (INIS)

Castiblanco, L.A.; Sarta, J.A.

1997-01-01

The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

International Nuclear Information System (INIS)

Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

2014-10-01

Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

International Nuclear Information System (INIS)

Ricardo, Carla Pereira; Mesquita, Amir Zacarias

2007-01-01

During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

Reactor Structure Materials: Corrosion of Reactor Core Internals

International Nuclear Information System (INIS)

Van Dyck, S.

2000-01-01

The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

Application of safety checklist to the analysis of the IEA-R1 reactor water retreatment system

International Nuclear Information System (INIS)

Sauer, Maria Eugenia Lago Jacques; Sara Neto, Antonio Jorge; Lima, Toni Carlos Caboclo de; Ribeiro, Maria Alice Morato

2005-01-01

In 1999, the management of the IEA-R1 Research Reactor (pool type - 5 MWth), located at IPEN/CNEN-SP, started the evaluation of the Reactor Pool Water Retreatment System to identify operational aspects, which could compromise the operators safety. The purpose was to identify and propose enhancements to the system which would be installed to substitute for the existing one. This process was conducted through a qualitative study of the system in operation. This study was carried out by a team composed of specialists in reactor operation, systems maintenance and radiological protection, and one safety analyst. The study consisted, basically, in local inspections to verify the physical and operational conditions of each equipment / component as well as aspects related to maintenance activities of the system. The process control and the operator procedures associated with the retreatment of the reactor pool water were also reviewed. The methodology adopted to develop the study was based in process hazard analysis technique named Safety Checklist. This paper presents a summary of this study and the main results obtained. Some operational and safety problems identified, the prevention and/or correction means to avoid them, and the recommendations and suggestions that have been implemented to the new design of the IEA-R1 Reactor Water Retreatment System, whose installation was concluded in 2003, are also presented. (author)

[Enhanced electro-chemical oxidation of Acid Red 3R solution with phosphotungstic acid supported on gamma-Al2O3].

Science.gov (United States)

Yue, Lin; Wang, Kai-Hong; Guo, Jian-Bo; Yang, Jing-Liang; Liu, Bao-You; Lian, Jing; Wang, Tao

2013-03-01

Supported phosphotungstic acid catalysts on gamma-Al2O3 (HPW/gamma-Al2O3) were prepared by solution impregnation and characterized by FTIR, XRD, TG-DTA and SEM. The heteropolyanion shows a Keggin structure. Electro-chemical oxidation of Acid Red 3R was investigated in the presence of HPW supported on gamma-Al2O3 as packing materials in the reactor. The results show that HPW/gamma-Al2O3 has a good catalytic activity for decolorization of Acid Red 3R. When HPW loading was 4.6%, pH value of Acid Red 3R was 3, the voltage was 25.0 V, air-flow was 0.04 m3 x h(-1), and electrode span was 3.0 cm, the decolorization efficiency of Acid Red 3R can reach 97.6%. The removal rate of color had still about 80% in this electro-chemical oxidation system, after HPW/gamma-Al2O3 was used for 10 times, but active component loss existed. The interim product was analyzed by means of Vis-UV absorption spectrum. It shows that the conjugated structure of dye is destroyed primarily.

The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

1999-01-01

The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

Study on effects of development of reactor constant in fast reactor analysis

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

Science.gov (United States)

2012-02-13

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-264; NRC-2012-0026] Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108 AGENCY: Nuclear Regulatory Commission... Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow Chemical Company...

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

International Nuclear Information System (INIS)

Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

2009-01-01

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.

1989-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

International Nuclear Information System (INIS)

Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

2013-01-01

The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

Science.gov (United States)

2013-01-28

... University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating License No. R-115, for the University of Illinois... Operating License No. R-115 is terminated. The above referenced documents may be examined, and/or copied for...

International R and D project on development of coated particle fuel for innovative reactors

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

The paper presents an outline for an international collaborative project of coated particle fuel development for innovative reactors. Specific issues include identification of R and D needs and the Member State facilities for meeting the needs followed by development and demonstration of technology. (author)

The evolution of doses in the IEA-R1 reactor environment and tendencies based on the current results

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio

2016-01-01

The IPEN / CNEN-SP have a Nuclear Research Reactor-NRR named IEA-R1, in operation from 1957. It is an open swimming pool reactor using light water as shielding, moderator and as cooling, the volume of this pool is 273m 3 .Until 1995 the reactor operated daily at a power of 2,0 MW. From June of that year, after a few safety modifications the reactor began operating in continuous way from Monday to Wednesday without shutdown totalizing 64 hours per week and the power was increased to 4,5MW also. Because of these changes, continuous operation and increased power, workers' doses would tend to increase. In the past several studies were conducted seeking ways to reduce the workers' doses. A study was made on the possibility to introduce a shielding at the top of the reactor core with a hot water layer. Studies have shown that a major limitation for operating a reactor at high power comes from the gamma radiation emitted by the sodium-24. Other elements such as magnesium-27, aluminum-28, Argon-51, contribute considerably to the water activity of the pool. The introduction of a hot water layer on the swimming pool would form a layer of surface, stable and free of radioactive elements with a 1.5m to 2m thickness creates a shielding to radiation from radioactive elements dissolved in water. Optimization studies proved that the installation of the hot layer was not necessary for the regime and the current power reactor operation, because other procedures adopted were more effective. From this decision the Radiological Protection Reactor Team, set up a dose assessment program to ensure them remained in low values based on principles established in national and international standards. The purpose of this paper is to analyze the individual doses of OEI (Occupationally Exposed Individual), which will be checked increasing doses resulting from recent changes in reactor operation regime and suggested viable safety and protection options, in the first instance to reducing

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

International Nuclear Information System (INIS)

Bittelli, U.D.

1988-01-01

This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

Total synthesis of (3S, 5R, 3'S, 5'R)-capsorubin

International Nuclear Information System (INIS)

Frederico, Daniel; Constantino, Mauricio G.; Donate, Paulo M.

2009-01-01

The total synthesis of enantiomerically enriched (3S, 5R, 3'S, 5'R)-capsorubin (1) by aldol condensation of (1R, 4S)-1-(4-hydroxy-1,2,2-trimethyl-cyclopentyl)ethanone (2a) and crocetindial (3) is described. An alternative, short eight-step synthesis of the optically active compound 2a (ee 89%) is also reported. (author)

Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

International Nuclear Information System (INIS)

Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

1987-01-01

The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

Experience with the RE fuel transition at the Studsvik R2 reactor

International Nuclear Information System (INIS)

Pazsit, I.; Saltvedt, K.

1991-01-01

Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

R and D for back-end options for irradiated research reactor fuel in Germany

International Nuclear Information System (INIS)

Bruecher, H.; Curtius, H.; Fachinger, J.

2001-01-01

Out of 11.5 t of irradiated fuel arising from German research reactors until the end of this decade, 3.9 t are intended to be returned to the USA, and 2.3 t are expected to be recycled for reuse of uranium. The remaining 5.3 t, as well as the fuel irradiated after the year 2010, will have to follow the domestic back-end option of extended dry interim storage in Castor-type casks, followed by disposal in a deep geological repository. R and D is going on in the Research Centre Juelich to investigate the long-term behaviour of U-Al based fuel in a salt repository. First results from leaching experiments show I) a fast dissolution of the fuel with mobilization of its radionuclide inventory, and 2) the following formation of amorphous Al-Mg-hydroxide phases. Long-lived actinides from the fuel were shown to be fixed in these phases and hence immobilized. Future R and D will be to investigate the nature and stability of these phases for long-term safety assessments. Investigations will have to be extended to cover alternative disposal sites (granite clay) as well as different (e.g. silicon based) fuels. (author)

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

Deliberation of Post 3.11 Fast Reactor R&D Strategy in Japan

International Nuclear Information System (INIS)

Kondo, Shunsuke

2013-01-01

• New nuclear energy strategy is still in the process of deliberation in Japan. Though many of idled plants will be restarted after renovation of their safety features in accordance with new safety rules set by the NRA, the contribution of nuclear power in Japan will probably not return to the level before 3.11. • Japan is in the process of reviewing its strategy for SFR R&D with a view to making it compatible with the new situation to be realized under new safety regulation as well as new energy strategy to be formulated within a year. • It is contemplated that major emphasis of the SFR R&D should be on a) the completion and use of MONJU, b) FR safety and c) waste minimization, not to mention the making effective use of our knowledge basis and research and engineering capabilities cultivated through the FaCT project in the development of sustainable nuclear energy systems worldwide for future generations of mankind. • Promotion of international cooperation should be an essential ingredient of the strategy

Study of neutronic flux in IPR-R1 reactor with MCNPX; Estudo do fluxo neutronico no reator IPR-R1 com o MCNPX

Energy Technology Data Exchange (ETDEWEB)

Melo, J.A.S.; Castrillo, L.S., E-mail: julio.angelo@poli.br, E-mail: lazara@poli.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica; Oliveira, R.M.B.M., E-mail: romero.matias@educacao.pe.gov.br [Secretaria Executiva de Educacao do Estado de Pernambuco (SEE), Recife, PE (Brazil)

2016-11-01

MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutron flux of 7% in the central tube and 5% on the turntable.

Biological biogas upgrading capacity of a hydrogenotrophic community in a trickle-bed reactor

International Nuclear Information System (INIS)

Rachbauer, Lydia; Voitl, Gregor; Bochmann, Günther; Fuchs, Werner

2016-01-01

Highlights: • Data on long term operation of a system supplied with real biogas are presented. • Ex-situ biological methanation is feasible for biogas upgrading. • Gas quality obtained complies with strictest direct grid injection criteria. • Biomethane can act as flexible storage for renewable surplus electricity. - Abstract: The current study reports on biological biogas upgrading by means of hydrogen addition to obtain biomethane. A mesophilic (37 °C) 0.058 m"3 trickle-bed reactor with an immobilized hydrogenotrophic enrichment culture was operated for a period of 8 months using a substrate mix of molecular hydrogen (H_2) and biogas (36–42% CO_2). Complete CO_2 conversion (> 96%) was achieved up to a H_2 loading rate of 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d, corresponding to 2.3 h gas retention time. The optimum H_2/CO_2 ratio was determined to be between 3.67 and 4.15. CH_4 concentrations above 96% were achieved with less than 0.1% residual H_2. This gas quality complies even with tightest standards for grid injection without the need for additional CO_2 removal. If less rigid standards must be fulfilled H_2 loading rates can be almost doubled (10.95 versus 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d) making the process even more attractive. At this H_2 loading the achieved methane productivity was 2.52 m_n"3 CH_4/m"3_r_e_a_c_t_o_r _v_o_l_. × d. In terms of biogas this corresponds to an upgrading capacity of 6.9 m_n"3 biogas/m"3_r_e_a_c_t_o_r _v_o_l_. × d. The conducted experiments demonstrate that biological methanation in an external reactor is well feasible for biogas upgrading under the prerequisite that an adequate H_2 source is available.

Investigation of MOZART experimental data and analysis of MOZART experiment using JFS-3-J3.2R group constant

International Nuclear Information System (INIS)

Kaise, Yoichiro; Osada, Hiroo

2003-03-01

Various critical experiments have been analyzed and evaluated in Japan Nuclear Cycle Development Institute (JNC) to improve the accuracy of prediction for nuclear characteristics of fast breeder reactors. This report describes update of the analysis of Monju Zebra Assembly Reactor Test (MOZART) reflecting a recent development of JNC analysis scheme. The main results are as follows: (1) Compilation of spectrum measurements: Spectrum measurement data are newly compiled including energy structure and geometrical information. (2) Reevaluation of atomic number density data: Atomic number density data were reevaluated considering impurities that had been neglected in the past analysis and reflecting a JNC standard analysis scheme. The revision of the data successfully reduces core type dependence of C/E values for criticality from 0.4%dk to 0.1%dk. (3) Analyses using JFS-3-J3.2R group constant set: The base-calculation and correction factors were fully reevaluated suing JFS-3-J3.2R group constant set and the results were compared with those using JFS-3-J3.2. For criticality, C/E values become smaller by 0.1%dk, which tendency is consistent with that observed in the analysis of JUPITER experiment. Reduction of B-10 concentration dependence from 7% to 1% is observed in C/E values for control rod worth, and 10% improvement are for Na void reactivity. These improvements are attribute to the revision of the group constant set and analysis scheme. The correction factors are confirmed to be insensitive to the revision of group constant sets. (author)

Removal in a lump of JRR-3 nuclear reactor

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

1989-01-01

The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Spatial distribution of the neutron flux in the IEA-R1 reactor core obtained by means of foil activation

International Nuclear Information System (INIS)

Mestnik Filho, J.

1979-01-01

A three-dimensional distribution of the neutron flux in IEA-R1 reactor, obtained by activating gold foils, is presented. The foils of diameter 8mm and thickness 0,013mm were mounted on lucite plates and located between the fuel element plates. Foil activities were measured using a 3x3 inches Nal(Tl) scintilation detector calibrated against a 4πβγ coincidence detector. Foil positions were chosen to minimize the errors of measurement; the overall estimated error on the measured flux is 5%. (Author) [pt

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Radio-activity measurements inside the pressure-vessel of the reactor G 3 after 4 years operation

International Nuclear Information System (INIS)

Chassany, J.Ph.; Guillermin, P.; Delmar, J.

1965-01-01

At the end of the piping coming into the vessel, the dose rate reached 75 mR/hr and 100 mR/hr near the deflector. On the other side of this deflector it was still 100 mR/hr and then increased rapidly to over 1 R/hr at 1 metre distance from the starting-up chambers. On the sides, the flux tended to decrease (80 mR/hr) and was 2 R/hr at a height of 3 metres. This dose rate could certainly have been decreased by discharging the peripheral zone of the reactor. Consequently it should be possible to intervene if necessary, on condition that great care is taken to avoid contamination and that the total dose is followed as precisely as possible during the operations. (authors) [fr

Energy transfer in LaF3: R3+, Pr3+ (where R = Nd, Dy)

International Nuclear Information System (INIS)

Reddy, B.R.; Venkateswarlu, P.

1982-01-01

Fluorescence is observed for 1 D 2 levels of Pr 3+ on exciting the higher lying level 3 P 0 in doubly doped systems LaF 3 : R 3+ , Pr 3+ (R = Nd or Dy) but not in LaF 3 :Pr 3+ . From the recorded excitation spectra, and the measured decay times, it has been found that the drain mechanism of population to 1 D 2 levels is caused in doubly doped systems by ion-pair relaxation between Pr 3+ ( 3 P 0 -- 1 D 2 ) and Nd 3+ (Z-X) or Dy 3+ (Z-Y) ions

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

Severe Accident R and D for Enhanced CANDU-6 Reactors

International Nuclear Information System (INIS)

Nitheanandan, Thambiayah

2012-01-01

CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

Energy Technology Data Exchange (ETDEWEB)

Klann, R. T.; Perret, G.; Nuclear Engineering Division

2007-10-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

Annual report of department of research reactor, 1992

International Nuclear Information System (INIS)

1993-12-01

The department of research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1992 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as well as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Calculation of neutronic parameters of IEA-R1 reactor and purpose of a new configuration

International Nuclear Information System (INIS)

Kosaka, N.; Fanaro, L.C.C.B.; Yamaguchi, M.

1989-01-01

The program for reducing the fuel enrichment of the IEA-R1 reactor considers fuel plates containing U308-AL with 19,9% of U-235. The geometry of the new 18 fuel plate fuel elements has been kept the same. This work describes the calculation methods utilized at IPEN-CNEN/SP and some neutronic parameters of the present configuration of IEA-R1 as well as for a new configuration porposed with a new LEU fuel element are shown. (author) [pt

Operating Experience at the Aagesta Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Sandstroem, S [ed.

1966-09-15

Sweden's first nuclear power reactor Agesta, achieved criticality on July 17, 1963. Full power (65 MW{sub t}) was attained on March 20, 1964. Aagesta is a heavy water cooled and moderated pressure vessel reactor used for production of electricity as well as for district heating. The design, assembly and construction etc, of the reactor was described in detail in a staff report by AB Atomenergi, 'The Aagesta Nuclear Power Station' edited by B McHugh, which was published in September, 1964. In the book experiences from the commissioning and the first operation of the reactor were reported as well as findings from the extensive reactor physics studies made during this period. The report now presented is written by members of the operating team at Aagesta since its start. It reflects in general the experiences up to the end of 1965. The Aagesta Log, however, covers the period up to the normal summer stop 1966. The reactor has hitherto produced 506,000 MWh power of which 48,700 MWh have been electric power. In July 1965 the responsibility for the reactor operation was taken over by the Swedish State Power Board from AB Atomenergi, which company had started the reactor and operated it until the summer break 1965.

Operating Experience at the Aagesta Nuclear Power Station

International Nuclear Information System (INIS)

Sandstroem, S.

1966-09-01

Sweden's first nuclear power reactor Agesta, achieved criticality on July 17, 1963. Full power (65 MW t ) was attained on March 20, 1964. Aagesta is a heavy water cooled and moderated pressure vessel reactor used for production of electricity as well as for district heating. The design, assembly and construction etc, of the reactor was described in detail in a staff report by AB Atomenergi, 'The Aagesta Nuclear Power Station' edited by B McHugh, which was published in September, 1964. In the book experiences from the commissioning and the first operation of the reactor were reported as well as findings from the extensive reactor physics studies made during this period. The report now presented is written by members of the operating team at Aagesta since its start. It reflects in general the experiences up to the end of 1965. The Aagesta Log, however, covers the period up to the normal summer stop 1966. The reactor has hitherto produced 506,000 MWh power of which 48,700 MWh have been electric power. In July 1965 the responsibility for the reactor operation was taken over by the Swedish State Power Board from AB Atomenergi, which company had started the reactor and operated it until the summer break 1965

Operating Experience at the Aagesta Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Sandstroem, S. (ed.)

1966-09-15

Sweden's first nuclear power reactor Agesta, achieved criticality on July 17, 1963. Full power (65 MW{sub t}) was attained on March 20, 1964. Aagesta is a heavy water cooled and moderated pressure vessel reactor used for production of electricity as well as for district heating. The design, assembly and construction etc, of the reactor was described in detail in a staff report by AB Atomenergi, 'The Aagesta Nuclear Power Station' edited by B McHugh, which was published in September, 1964. In the book experiences from the commissioning and the first operation of the reactor were reported as well as findings from the extensive reactor physics studies made during this period. The report now presented is written by members of the operating team at Aagesta since its start. It reflects in general the experiences up to the end of 1965. The Aagesta Log, however, covers the period up to the normal summer stop 1966. The reactor has hitherto produced 506,000 MWh power of which 48,700 MWh have been electric power. In July 1965 the responsibility for the reactor operation was taken over by the Swedish State Power Board from AB Atomenergi, which company had started the reactor and operated it until the summer break 1965.

Synthesis of carbasugars from aldonolactones, part III - A study on the allylic substitution of (1R,5R,8R)- and (1R,5R,8S)-8-hydroxy-2-oxabicyclo[3.3.0]oct-6-en-3-one derivatives - Preparation of (1S,2R,3R)-9-[2-hydroxy-3-(2-hydroxyethyl)cyclopent-4-en-1-yl]-9H-adenine

DEFF Research Database (Denmark)

Johansen, Steen Karsk; Lundt, Inge

2001-01-01

The palladium-catalyzed substitution of acylated (1R,5R,8R)- and (1R,SR,8S)-8-hydroxy-2-oxabicyclo[3.3.0] ones has been studied using a number of C- and N-nucleophiles, In all cases, the exo derivatives (8R) were found to be more reactive than the corresponding endo derivatives (8S). The reaction...... with these nucleophiles. Additionally, Mitsunobu substitution of (1R,5R,8R)-8-hydroxy-2-oxabicyclo[3.3.0]oct-B-en-3-one (3) with 6-chloropurine, followed by reduction of the lactone moiety and treatment with Liquid ammonia, gave the carbocyclic nucleoside (1S,2R,3R)-9-[2-hydroxy-3-(2-hydroxyethyl)cyclopent-4-en-1-yl]-9H...

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

International Nuclear Information System (INIS)

Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

2008-01-01

The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

Status of French R and D for advanced light water reactors

International Nuclear Information System (INIS)

Nigon, J.L.

1987-01-01

Present PWRs lead to a significant reduction of electricity cost when compared to other sources. Then it seems reasonable to keep the main features of PWRs when looking for improvements of investment cost, of operating and fuel costs, of flexibility and of safety. Besides that we have to think about uranium conservation; if nuclear starts again in many countries, as we hope, the uranium market could get into a crisis during the first half of the 21st century, and uranium shortage could become a reality. Advanced PWRs are also aimed at fissile material saving. The three French partners CEA, EdF and FRAMATOME decided to lead a three year programme 1984-1987. FRAMATOME in fact had started a little bit earlier, namely in 1982 (first publication in RNG - a French technical journal) and developed the RCVS, spectral shift convertible reactor core (for both Uranium and Plutonium fuel). FRAMATOME's effort is estimated to about 40.10 6 FF per year. EdF, R and D Division, is associated with this feasibility study. CEA performs an R and D programme, the objectives of which are: to support FRAMATOME for RCVS design; to explore a wider range of parameters in order to estimate the feasibility and the interest of tight lattice PWR cores. Simultaneously, EdF is defining the preliminary specifications of ''REP 2000'' (future standard for French PWRs in the year 2000 and following); the objectives of REP 2000 are: load follow capacity; cost effectiveness; operation flexibility. FRAMATOME's RCVS and the CEA RSM feasibility study have to be considered in this context. The main objectives are: 1) To improve performances, safety and to minimise cost; 2) To save fissile materials according to a global strategy; 3) While minimum modifications of present PWR components will be accepted. The R and D budget for PWRs (outside safety) of French CEA is around 450 10 6 FF per year. Among this, 40 10 6 FF/year are devoted to tight lattice core feasibility studies (period 1984-1987). 3 figs

Device for neutron flux monitoring in IEA-R1 reactor using rhodium self powered neutron detector; Dispositivo de mapeamento de fluxo de neutron atraves do SPN/Rodio no IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Ricci Filho, Walter; Fernando, Alberto de Jesus; Jerez, Rogerio; Tondin, Julio B.M.; Pasqualetto, Hertz [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

2000-07-01

The IEA-R1 reactor has undergone a modernization tio increase its operating power to 5 MW, in order to allow a more efficient production of radioisotopes. The objective of this work is to provide the reactor with flux monitoring device using a rhodium self powered neutron detector. Self powered detectors are rugged miniature devices with are increasingly being used for fixed in core reactor monitoring both for safety purposes and flux mapping. The work presents the results obtained with Rhodium-SPND in several irradiation position inside the reactor core. (author)

Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

International Nuclear Information System (INIS)

Torres, Walmir Maximo; Baptista Filho, Benedito Dias

2000-01-01

This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

U.S. Status of Fast Reactor Research and Technology

International Nuclear Information System (INIS)

Hill, Robert

2012-01-01

Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

Analysis of the IEA-R1 reactor start-up procedures - an application of the HazOp method

International Nuclear Information System (INIS)

Sauer, Maria Eugenia Lago Jacques

2000-01-01

An analysis of technological catastrophic events that took place in this century shows that human failure and vulnerability of risk management programs are the main causes for the occurrence of accidents. As an example, plants and complex systems where the interface man-machine is close, the frequency of failures tends to be higher. Thus, a comprehensive knowledge of how a specific process can be potentially hazardous is a sine qua non condition to the operators training, as well as to define and implement more efficient plans for loss prevention and risk management. A study of the IEA-R1 research reactor start-up procedures was carried out, based upon the methodology Hazard and Operability Study (HazOp). The analytical and qualitative multidisciplinary HazOp approach provided means to a comprehensive review of the reactor start-up procedures, contributing to improve the understanding of the potential hazards associated to deviations on performing this routine. The present work includes a historical summary and a detailed description of the HazOp technique, as well as case studies in the process industries and the use of expert systems in the application of the method. An analysis of 53 activities of the IEA-R1 reactor start-up procedures was made, resulting in 25 recommendations of changes covering aspects of the project, operation and safety of the reactor. Eleven recommendations have been implemented. (author)

Plan and reports of coupled irradiation (JRR-3 and JOYO of research reactors) and hot facilities work (WASTEF, JMTR-HL, MMF and FMF). R and D project on irradiation damage management technology for structural materials of long-life nuclear plant

International Nuclear Information System (INIS)

Matsui, Yoshinori; Yamamoto, Masaya; Yoshitake, Tsunemitsu; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ichikawa, Shoichi; Yamagata, Ichiro; Soga, Tomonori; Yonekawa, Minoru; Kitamura, Ryoichi; Miyake, Osamu; Takahashi, Hiroyuki; Ishikawa, Kazuyoshi; Kikuchi, Taiji; Usami, Koji; Endo, Shinya; Ichise, Kenichi; Numata, Masami; Onozawa, Atsushi; Aizawa, Masao; Kusunoki, Tsuyoshi; Nakata, Masahito; Abe, Kazuyuki; Ito, Kazuhiro; Takaya, Shigeru; Nagae, Yuji; Wakai, Eiichi; Aoto, Kazumi

2010-03-01

'R and D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant' was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of 'Evaluation of Irradiation Damage Indicator' in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research and Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency. (author)

Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

Energy Technology Data Exchange (ETDEWEB)

2009-01-15

This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

Decommissioning of the Nuclear Reactors R2 and R2-0 at Studsvik, Sweden. General Data as called for under Article 37 of the Euratom Treaty

International Nuclear Information System (INIS)

2009-01-01

This document describes the plans for decommissioning of the nuclear research and material test reactors R2 and R2-0, situated at the Studsvik site close to the city of Nykoeping, Sweden. The purpose of the document is to serve as information for the European Commission, and to fulfil the requirements of Article 37 of the Euratom Treaty. Studsvik is situated on the Baltic coast, about 20 km east of Nykoeping and 80 km southwest of Stockholm. The site comprises the reactors R2 and R2-0 and several facilities for material investigation and radioactive waste treatment and storage. The reactors were used for a number of different purposes from 1960 until June 2005, when they were shut down following a decision by the operator. Decommissioning of the reactor facility is planned to be completed in 2016 after dismantling and conditioning of radioactive parts and demolition of the facility. Solid and liquid radioactive wastes from the dismantling activities will be treated and stored on-site awaiting final disposal. The waste treatment facilities, which are situated in other buildings at the Studsvik site, are planned to continue operation during and after the decommissioning of the reactor facility. All nuclear fuel has been transferred to a separate storage facility and is being shipped to the US according to existing agreements. The objective of the planned dismantling activities is to achieve clearance of the facility to make it possible to either demolish the buildings or use them for other purposes. The operator has divided the planning for dismantling and demolition of the facility into three phases [1]: Dismantling 1, including primary system decontamination, dismantling of the reactors with systems in the reactor pool, draining, cleaning and temporary covering of the reactor pool. This phase has begun and is due to last till approximately December 2009. Dismantling 2, including dismantling of systems in the reactor facility, removal of equipment, radiological

Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

Energy Technology Data Exchange (ETDEWEB)

Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

Neutron flux of 100kW in the irradiation terminals of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante Marco

2009-01-01

In this work, it was carried out a study of the neutron flux in the IPR-R1 TRIGA reactor irradiation facilities: rotary specimen rack (RSR), pneumatic transfer tube two (PTT2) and the central thimble (CT). The objective was to obtain the neutron flux profile on the RSR, which has forty irradiation positions, and also values for the thermal and epithermal neutron fluxes of some RSR positions and also of the PTT2 and of the CT facility. It was applied the neutron activation analysis of a reference material, Al-Au (0.1%) alloy. Irradiations were performed on 16 different dates. It was concluded that for the RSR, the average value of thermal and epithermal neutron fluxes depends on the vertical position of the reactor control rods. Neutron flux variations along the RSR form a characteristic profile, whose values depend on the location of the irradiation position in the reactor core and on the control rods vertical position. In the RSR, the obtained values of thermal and epithermal neutron flux were (8.1 +- 0.3) x 10 11 n.cm -2 .s -1 , and (3.4 +- 0.2)x10 10 n.cm -2 .s -1 , respectively. For the PTT2 and the CT, the values for the epithermal neutron flux were respectively (3.3 +- 0.2) x 10 9 n.cm -2 .s -1 and (2.6 +- 0.1) x 10 11 n.cm -2 .s -1 . For these facilities, the thermal neutron flux was estimated, and the obtained values were (2.4 +- 0.2) x 10 11 n.cm -2 .s -1 and (2.8 +- 0.1)x10 12 n.cm -2 .s -1 for the PTT2 and the CT, respectively. (author)

Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

2001-01-01

The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

Considerations about decommissioning of the IEA-R1 research reactor and the future of its installations after shutdown

International Nuclear Information System (INIS)

Frajndlich, Roberto

2014-01-01

The IEA-R1 Nuclear Research Reactor, in operation since 1957, in the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), is one of the oldest research reactors in the world. However at some point in time in the future, as example of the other reactors, it will be shutdown definitively. Before that time actually arrives, the operational organization needs to plan the future of its installations and define the final destination of equipment and radioactive as well as non-radioactive material contained inside the installations. These and other questions should be addressed in the so called Preliminary decommissioning plan of the installation, which is the subject of this work. The work initially presents an over view about the theme and defines the general and specific objectives describing, in succession, the directions that the operating organization should consider for the formulation of a decommissioning plan. The present structure of the Brazilian nuclear sector emphasizing principally the norms utilized in the management of radioactive waste is also presented. A description of principle equipment of the IEA-R1 reactor which constitutes its inventory of radioactive and non-radioactive material is given. The work emphasizes the experience of the reactor technicians, acquired during several reforms and modifications of the reactor installations realized during its useful life time. This experience may be of great help for the decommissioning in the future. An experiment using the high resolution gamma spectrometric method and computer calculation using Monte Carlo theory were performed with the objective of obtaining an estimate of the radioactive waste produced from dismantling of the reactor pool walls. The cost of reactor decommissioning for different choices of strategies was determined using the CERREX code. Finally, a discussion about different strategies is presented. On the basis of these discussions it is concluded that the most advantageous

Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

International Nuclear Information System (INIS)

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

2013-01-01

The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

Science.gov (United States)

Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

2017-12-01

The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

D-3He fueled FRC reactor 'ARTEMIS-L'

International Nuclear Information System (INIS)

Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

1992-09-01

A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

Description of the RA-3 research reactor as a model facility

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

2011-07-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

International Nuclear Information System (INIS)

Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

2011-01-01

The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

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)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

International Nuclear Information System (INIS)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

2017-01-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

2017-11-01

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

R[ionuclide transport after the Chernobyl reactor accident and derivation of r[ioecological parameters

International Nuclear Information System (INIS)

Bonka, H.

1998-01-01

Since due to the nuclear reactor accident in Chernobyl r[ionuclides arrived in the vicinity of Aachen, the enhancement of the local dose rate, the deposition of the different r[ionuclides on ground and vegetation and the transport of the r[ionuclides into the environment were measured. Partly the measurements were continued until today. Very informative time sequences of the specific activity in grass, food, cow's milk, beef, in the different plants, trees, ploughed soil and undisturbed soil, mushrooms, game, in humans etc. resulted. During different private and official journeys in the old Laender of the Federal Republic of Germany surface covering measurements of the 134 Cs and 137 Cs activity deposited on grass land at different places were carried out. These data were implemented into a map on ground contamination in 1986 in Germany, published in 1991 by the Institute for Water, Soil and Air Hygiene of the Federal Public Health Department in Berlin. Transfer factors soil-grass were measured in the whole Federal Republic of Germany analyzing grass samples which were partly taken at the same time. A large amount of r[ioecological parameters could be derived from the different time sequences. These are in particular: The deposition velocity for iodine and particle bound r[ioanuclides on grass and in forests, the rainout coefficient in dependence of the precipitation intensity, the retention factors on grass, the biological half-life time on grass, the transfer factor soil-grass in dependence of time, the transfer factor food-milk during the pasture period and during stable stay, the transfer factor food-beef, the transfer factors in eatable mushrooms, the translocation factor of cesium in cereals etc. A multi-compartment model was developed to calculate the specific Cs activity in cow's milk and beef. The specific activity in milk can be calculated sufficiently exact using a simple single compartment model. The correlation of the specific Cs activity in spruce

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

International Nuclear Information System (INIS)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-01

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

2008-01-15

This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

Resolution of 1-n-butyl-3-methyl-3-phospholene 1-oxide with TADDOL derivatives and calcium salts of O,O'-Dibenzoyl-(2R,3R)- or O,O'-di-p-toluoyl-(2R,3R)-tartaric acid.

Science.gov (United States)

Bagi, Péter; Fekete, András; Kállay, Mihály; Hessz, Dóra; Kubinyi, Miklós; Holczbauer, Tamás; Czugler, Mátyás; Fogassy, Elemér; Keglevich, György

2014-03-01

The resolution methods applying (-)-(4R,5R)-4,5-bis(diphenylhydroxymethyl)-2,2-dimethyldioxolane ("TADDOL"), (-)-(2R,3R)-α,α,α',α'-tetraphenyl-1,4-dioxaspiro[4.5]decan-2,3-dimethanol ("spiro-TADDOL"), as well as the acidic and neutral Ca(2+) salts of (-)-O,O'-dibenzoyl- and (-)-O,O'-di-p-toluoyl-(2R,3R)-tartaric acid were extended for the preparation of 1-n-butyl-3-methyl-3-phospholene 1-oxide in optically active form. In one case, the intermediate diastereomeric complex could be identified by single-crystal X-ray analysis. The absolute P-configuration of the enantiomers of the phospholene oxide was also determined by comparing the experimentally obtained and calculated CD spectra. © 2014 Wiley Periodicals, Inc.

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

International Nuclear Information System (INIS)

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

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.)

Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

International Nuclear Information System (INIS)

Xolocostli M, J.V.; Gomez T, A.M.; Palacios H, J.C.

2006-01-01

The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor; Estudo da viabilidade de aplicação do método Prompt Gamma Neutron Activation Analysis (PGNAA) no reator TRIGA IPR-R1

Energy Technology Data Exchange (ETDEWEB)

Guerra, Bruno Teixeira

2016-07-01

The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

The R and D issues necessary to achieve the safety design of commercialized liquid-metal cooled fast reactors

International Nuclear Information System (INIS)

Shoji, Kotake; Koji, Dozaki; Shigenobu, Kubo; Yoshio, Shimakawa; Hajime, Niwa; Masakazu, Ichimiya

2002-01-01

Within the framework of the feasibility study on commercialized fast reactor cycle systems (hereafter described as F/S), the safety design principle is investigated and several kinds of design studies are now in progress. Among the designs for liquid-metal cooled fast reactor (LMR), the advanced loop type sodium cooled fast reactor (FR) is one of the promising candidate as future commercialized LMR. In this paper, the safety related research and development (R and D) issues necessary to achieve the safety design are described along the defence-in-depth principle, taking account of not only the system characteristics of the advanced loop concepts but also design studies and R and D experiences so far. Safety issues related to the hypothetical core disruptive accidents (CDA) are emphasized both from the prevention and mitigation. A re-criticality free core concept with a special fuel assembly is pursued by performing both analytical and experimental efforts, in order to realize the rational design and to establish easy-to-understand safety logic. Sodium related issues are also given to ensure plant availability and to enhance the acceptability to the public. (authors)

Nuclear research reactor IEA-R1 heat exchanger inlet nozzle flow - a preliminary study

International Nuclear Information System (INIS)

Angelo, Gabriel; Andrade, Delvonei Alves de; Fainer, Gerson; Angelo, Edvaldo

2009-01-01

As a computational fluid mechanics training task, a preliminary model was developed. ANSYS-CFX R code was used in order to study the flow at the inlet nozzle of the heat exchanger of the primary circuit of the nuclear research reactor IEA-R1. The geometry of the inlet nozzle is basically compounded by a cylinder and two radial rings which are welded on the shell. When doing so there is an offset between the holes through the shell and the inlet nozzle. Since it is not standardized by TEMA, the inlet nozzle was chosen for a preliminary study of the flow. Results for the proposed model are presented and discussed. (author)

Water chemical control of the TRIGA IPR-R1 reactor primary cooling system

International Nuclear Information System (INIS)

Auler, Lucia M.L.A; Chaves, Renata D.A.; Palmieri, Helena E.L.; Menezes, Maria Angela de B.C.; Oliveira, Paulo F.; Kastner, Geraldo F.; Damazio, Ilza; Fagundes, Oliene dos R.; Cintra, Maria Olivia C.; Andrade, Geraldo V. de; Amaral, Angela M.; Franco, Milton B.; Fortes, Flavio; Gomes, Nilton Carlos; Vidal, Andrea; Maretti Junior, Fausto; Knupp, Eliana A.N.; Souza, Wagner de; Guedes, Joao B.; Furtado, Renato C.S.

2013-01-01

The TRIGA Mark I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. Is has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. The aim of this work was to check the cooling water quality from the pool reactor. A water sampling plan was proposed (May, 2011 - June, 2012) and presents the results obtained in this period. The natural radioactivity level as gross alpha and gross beta activity and other chemical parameters (pH and electric conductivity) of the samples were analyzed. Some instrumental techniques were used: potentiometric methods (pH), conductometric methods (electrical conductivity, EC) and gross α and gross β proportional counting system). (author)

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

Energy Technology Data Exchange (ETDEWEB)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H., E-mail: songm@dgu.edu [Department of Mechanical, Robotics, and Energy Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of)

2015-03-15

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

International Nuclear Information System (INIS)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

2015-01-01

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made

Preliminary analysis of control rod accidents in the CRCN-R1 multipurpose reactor core of Recife in Brazil

International Nuclear Information System (INIS)

Souza dos Santos, Rubens; Rubens Maiorino, Jose

1999-01-01

The paper shows some results of the neutronic accident analyses arisen by uncontrolled control rod withdrawal, based on the Conceptual Project of the CRCN-R1 MultiPurpose Reactor of Recife. In that reactor, a project of the CNEN/Brazil, under the leadership of the IPEN/Sao Paulo, is verified the thermal hydraulic limits in the reactor core during transients that simulate startup and power operation accidents. It has utilized a computer program that solved the kinetic equations based on multigroup diffusion theory, in our case we have used 4 energy groups, Two-Dimensional X-Y in the space, and 6 groups of delayed neutrons. A simple model of feedback is admitted in the capture and scattering macroscopic cross sections, in the fuel regions, temperature and coolant densities dependents. Based on those models, the results demonstrated that the reactor exhibits good degree of safety. (author)

Nuclear R and D program in Indonesia and selection of future research reactor to support it

International Nuclear Information System (INIS)

Baiquni, A.; Subki, I.

1981-01-01

The nuclear R and D program selection decision is described as a phased program, each phase having its specific objective. The elements of each phase are identified and related with the objective, from which the activities of each element are also broadly outlined. To support the nuclear R and D program and to realize the objectives in each phase, the research facilities are also developed. A new nuclear development center housing a multipurpose reactor (MPR) and various laboratories are also described. The choice of the MPR and its criteria are also described briefly

Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

International Nuclear Information System (INIS)

Mesquita, A.Z.; Maretti, F. Jr.; Rezende, H.C.; Tambourgi, E.B.

2004-01-01

The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

The French R and D programme for fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Auchapt, P.; Bourgeois, M.; Calame-Longjean, A.; Miquel, P.; Sauteron, J.

1979-01-01

The process employed is the Purex process adapted to the specific case of fast breeder reactor fuels. The results achieved have demonstrated that the aqueous method can be applied to these fuels: nearly ten years of operation in the ATl workshop which reprocesses RAPSODIE fuels, and the good results obtained at the Marcoule pilot facility on large batches of fuel attest to this achievement. The CEA effort continues principally on extrapolation to industrial scale, thanks mainly to experiments conducted on industrial prototypes and to the launching of the TOR project, which will, as of 1984, allow reprocessing of FBR fuels on a significant scale, and which will provide extensive additional resources for R and D activities

Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

CERN Document Server

Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

2000-01-01

Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

Presence of Tritium in the Cooling Circuits of the Reactors G2 and G3; Presence de tritium dans les circuits de refroidissement des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Estournel, R [Commissariat a l' Energie Atomique. Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

1962-07-01

In a reactor of the G 2-G 3 type, tritium can be formed by the neutronic bombardment of many elements present in the core. Tritium was found to be present in the cooling circuits of the reactors G 2 and G 3 in the water coming from the regeneration of the CO{sub 2} dehydrating columns. (author) [French] Dans un reacteur du type G 2 - G 3, le tritium peut etre forme par le bombardement. neutronique de nombreux elements existant dans le c r. La presence de tritium dans les circuits de refroidissement des reacteurs G 2 - G 3 a ete mis en evidence dans l'eau provenant de la regeneration des colonnes de deshydratation du CO{sub 2}. (auteur)

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

2016-09-01

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.; Chang, Y.W.

1990-01-01

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

Dose measurements in controlled area of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Alvarenga, F.L.; Junior, F.M.

2005-01-01

The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

Directory of Open Access Journals (Sweden)

Yeong-il Kim

2013-01-01

Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

Research and development of super light water reactors and super fast reactors in Japan

International Nuclear Information System (INIS)

Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

2011-01-01

Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

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.)

The ARCHER project (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D)

Energy Technology Data Exchange (ETDEWEB)

Knol, S., E-mail: knol@nrg.eu [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Fütterer, M.A. [Joint Research Centre, Institute for Energy, Petten (Netherlands); Roelofs, F. [Nuclear Research and consultancy Group (NRG), PO Box 25, NL-1755 ZG Petten (Netherlands); Kohtz, N. [TÜV Rheinland, Köln (Germany); Laurie, M. [Joint Research Centre, Institute for Transuranium elements, Karlsruhe (Germany); Buckthorpe, D. [UMAN, University of Manchester, Manchester (United Kingdom); Scheuermann, W. [IKE, Stuttgart University, Stuttgart (Germany)

2016-09-15

The European HTR R&D project ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) builds on a solid HTR technology foundation in Europe, established through former national UK and German HTR programs and in European framework programs. ARCHER runs from 2011 to 2015 and targets selected HTR R&D subjects that would specifically support demonstration, with a focus on experimental effort. In line with the R&D and deployment strategy of the European Sustainable Nuclear Energy Technology Platform (SNETP) ARCHER contributes to maintaining, strengthening and expanding the HTR knowledge base in Europe to lay the foundations for demonstration of nuclear cogeneration with HTR systems. The project consortium encompasses conventional and nuclear industry, utilities, Technical Support Organizations, R&D organizations and academia. ARCHER shares results with international partners in the Generation IV International Forum and collaborates directly with related projects in the US, China, Japan, the Republic of Korea and South Africa. The ARCHER project has finished, and the paper comprises an overview of the achievements of the project.

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

The past and the future in the forty years of the IPR-R1 TRIGA MARK I reactor operation

International Nuclear Information System (INIS)

Maretti Junior, Fausto

2008-01-01

Full text: The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. During these years a lot of irradiations, analysis , MSc and PhD thesis, training courses and isotopes production take place at the reactor. This paper describes the improvements made, the results obtained during the past 40 years, type of works realized, isotopes produced, the neutron activation analysis and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (authors)

The evolution of doses in the IEA-R1 reactor environment and tendencies based on the current results; Evolucao das doses no ambiente do Reator IEA-R1 e tendencias com base nos resultados atuais

Energy Technology Data Exchange (ETDEWEB)

Toyoda, Eduardo Yoshio

2016-11-01

The IPEN / CNEN-SP have a Nuclear Research Reactor-NRR named IEA-R1, in operation from 1957. It is an open swimming pool reactor using light water as shielding, moderator and as cooling, the volume of this pool is 273m{sup 3}.Until 1995 the reactor operated daily at a power of 2,0 MW. From June of that year, after a few safety modifications the reactor began operating in continuous way from Monday to Wednesday without shutdown totalizing 64 hours per week and the power was increased to 4,5MW also. Because of these changes, continuous operation and increased power, workers' doses would tend to increase. In the past several studies were conducted seeking ways to reduce the workers' doses. A study was made on the possibility to introduce a shielding at the top of the reactor core with a hot water layer. Studies have shown that a major limitation for operating a reactor at high power comes from the gamma radiation emitted by the sodium-24. Other elements such as magnesium-27, aluminum-28, Argon-51, contribute considerably to the water activity of the pool. The introduction of a hot water layer on the swimming pool would form a layer of surface, stable and free of radioactive elements with a 1.5m to 2m thickness creates a shielding to radiation from radioactive elements dissolved in water. Optimization studies proved that the installation of the hot layer was not necessary for the regime and the current power reactor operation, because other procedures adopted were more effective. From this decision the Radiological Protection Reactor Team, set up a dose assessment program to ensure them remained in low values based on principles established in national and international standards. The purpose of this paper is to analyze the individual doses of OEI (Occupationally Exposed Individual), which will be checked increasing doses resulting from recent changes in reactor operation regime and suggested viable safety and protection options, in the first instance to

US DOE Idaho national laboratory reactor decommissioning

International Nuclear Information System (INIS)

Szilagyi, Andrew

2012-01-01

The United States Department of Energy (DOE) primary contractor, CH2M-WG Idaho was awarded the cleanup and deactivation and decommissioning contract in May 2005 for the Idaho National Lab (INL). The scope of this work included dispositioning over 200 Facilities and 3 Reactors Complexes (Engineering Test Reactor (ETR), Materials Test Reactor (MTR) and Power Burst Facility (PBF) Reactor). Two additional reactors were added to the scope of the contract during the period of performance. The Zero Power Physics Reactor (ZPPR) disposition was added under a separate subcontractor with the INL lab contractor and the Experimental Breeder Reactor II (EBR-II) disposition was added through American Recovery and Reinvestment Act (ARRA) Funding. All of the reactors have been removed and disposed of with the exception of EBR-II which is scheduled for disposition approximately March of 2012. A brief synopsis of the 5 reactors is provided. For the purpose of this paper the ZPPR reactor due to its unique design as compared to the other four reactors, and the fact that is was relatively lightly contaminated and irradiated will not be discussed with the other four reactors. The ZPPR reactor was readily accessible and was a relatively non-complex removal as compared to the other reactors. Additionally the EBR-II reactor is currently undergoing D and D and will have limited mention in this paper. Prior to decommissioning the reactors, a risk based closure model was applied. This model exercised through the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Non-Time Critical Removal Action (NTCRA) Process which evaluated several options. The options included; No further action - maintain as is, long term stewardship and monitoring (mothball), entombment in place and reactor removal. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, beryllium, sodium (passivated and elemental), PCB oils and electrical components, lead

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

2014-01-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

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

Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

Science.gov (United States)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

Doping of monocrystalline silicon with phosphorus by means of neutron irradiation at the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carbonari, A.W.; Puget, M.A.C.

1990-11-01

The first neutron irradiation experiments with monocrystal silicon in the IEA-R1 research reactor of IPEN are related. The silicon is irradiated with phosphorus producing a N type semiconductor with a very small resistivity variation throughout the crystal volume. The neutrons induce nuclear reactions in Si-30 isotope and these atoms are then transformed in to phosphorous atoms. This process is known as Neutron Transmutation Doping. In order to irradiate the silicon crystals in the reactor, a specific device has been constructed, and it permits the irradiation of up to 2.5'' diameter monocrystals. (author)

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Reactor Structural Materials: Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chaouadi, R.

2000-01-01

The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

Feasibility studies of producing 99 Mo by capture in the IEA-R1 research reactor

International Nuclear Information System (INIS)

Concilio, Roberta; Mendonca, Arlindo Gilson; Maiorino, Jose Rubens

1998-01-01

Everyday the production of 99 Mo for 99m Tc generators, becomes more necessary, whose properties are ideal for medical diagnosis. This works presents a description and an analysis of the production of 99 Mo by radioactive capture at 98 Mo using the research reactor IEA-R1 in 5 MW and operating 5 days a week, referring to the use of targets, separation methods, total and specific activity attained and its limitations. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

Real-Time Monitoring of Neutron Capture Cross Section in the IPR-R1 TRIGA Research Reactor as a Fuel Temperature Function

Energy Technology Data Exchange (ETDEWEB)

Palma, D.A.P. [Comissao Nacional de Energia Nuclear, CNEN, General Severiano Street, 90, 22290-901, Rio de Janeiro (Brazil); Mesquita, A.Z.; Souza, R.M.G.P. [Comissao Nacional de Energia Nuclear, CNEN/CDTN, Av. Presidente Antonio Carlos, 6627, 31270-901, Belo Horizonte (Brazil); Martinez, A.S. [Programa de Engenharia Nuclear, COPPE/UFRJ, Av. Horacio Macedo, 2030, Bloco G, 21941- 914, Rio de Janeiro (Brazil)

2011-07-01

Nuclear reactor operators have to monitor the behaviour of different nuclear and design parameters that vary in time to ensure the operating safety of the reactor. In recent years several operating parameters for the IPR-R1 TRIGA research reactor were monitored and indicated in real-time by the data acquisition system developed for the reactor, with all the data being stored in a hard disk in the data acquisition computer, to build in this way a database. The goal of this work is to insert in the set of parameters already collected the neutron capture cross sections for the fuel, from the power and temperature numbers obtained in real-time. The experimental data was obtained by using a fuel element instrumented with temperature sensors, located in the core of the IPR-R1 TRIGA research reactor at the CDTN - Centre for Development of Nuclear. This information is useful for the continuous monitoring of the reaction rate in neutron capture. For that, a new analytical formulation is used for the Doppler broadening function proposed by Palma and Martinez which is free from special functions in its functional form and with easy computing implementation. The results obtained were satisfactory from the standpoint of accuracy in comparison with the numerical reference method and indicate that it is possible to carry out real-time monitoring of the neutron capture cross section in the fuel. (author)

(1R,2R,3R,4R,5S-2,3-Bis[(2S′-2-acetoxy-2-phenylacetoxy]-4-azido-1-[(2,4-dinitrophenylhydrazonomethyl]bicyclo[3.1.0]hexane

Directory of Open Access Journals (Sweden)

Robert McDonald

2008-02-01

Full Text Available In the title compound, C38H29N7O12, the five-membered ring adopts an envelope conformation in which the `flap' is cis to the cyclopropane group. This conformation is similar to those of other bicyclo[3.1.0]hexane analogues for which crystal structures have been reported. The absolute configuration of the stereogenic centers on the cyclopentane ring, as determined by comparison with the known configurations of the stereogenic centers in the (2S-2-acetoxy-2-phenylacetoxy groups, is 1(R, 2(R, 3(R, 4(R and 5(S. An intramolecular N—H...O hydrogen bond is present.

Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

International Nuclear Information System (INIS)

2015-01-01

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institute (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: 108m Ag, 110m Ag, 60 Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)

Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-07-01

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institute (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)

A review of fast reactor program in Japan

International Nuclear Information System (INIS)

1996-01-01

The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

Annual report of department of research reactor, 1999. April 1, 1999 - March 31, 2000

International Nuclear Information System (INIS)

2001-03-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3M (new JRR-3) and the JRR-4 and for the related R and D. Besides the decommissioning of the JRR-2 and RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 1999 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

R Reactor seepage basins soil moisture and resistivity field investigation using cone penetrometer technology, Savannah River Site, Aiken, South Carolina

International Nuclear Information System (INIS)

Harris, M.K.

2000-01-01

The focus of this report is the summer 1999 investigation of the shallow groundwater system using cone penetrometer technology characterization methods to determine if the water table is perched beneath the R Reactor Seepage Basins (RRSBs)

Neutrons characterization of the nuclear reactor Ian-R1 of Colombia

International Nuclear Information System (INIS)

Gonzalez P, L. X.; Martinez O, S. A.; Vega C, H. R.

2014-08-01

By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

Improved Dechlorinating Performance of Upflow Anaerobic Sludge Blanket Reactors by Incorporation of Dehalospirillum multivorans into Granular Sludge

Science.gov (United States)

Hörber, Christine; Christiansen, Nina; Arvin, Erik; Ahring, Birgitte K.

1998-01-01

Dechlorination of tetrachloroethene, also known as perchloroethylene (PCE), was investigated in an upflow anaerobic sludge blanket (UASB) reactor after incorporation of the strictly anaerobic, reductively dechlorinating bacterium Dehalospirillum multivorans into granular sludge. This reactor was compared to the reference 1 (R1) reactor, where the granules were autoclaved to remove all dechlorinating abilities before inoculation, and to the reference 2 (R2) reactor, containing only living granular sludge. All three reactors were fed mineral medium containing 3 to 57 μM PCE, 2 mM formate, and 0.5 mM acetate and were operated under sterile conditions. In the test reactor, an average of 93% (mole/mole) of the effluent chloroethenes was dichloroethene (DCE), compared to 99% (mole/mole) in the R1 reactor. The R2 reactor, with no inoculation, produced only trichloroethene (TCE), averaging 43% (mole/mole) of the effluent chloroethenes. No dechlorination of PCE was observed in an abiotic control consisting of sterile granules without inoculum. During continuous operation with stepwise-reduced hydraulic retention times (HRTs), both the test reactor and the R1 reactor showed conversion of PCE to DCE, even at HRTs much lower than the reciprocal maximum specific growth rate of D. multivorans, indicating that this bacterium was immobilized in the living and autoclaved granular sludge. In contrast, the R2 reactor, with no inoculation of D. multivorans, only converted PCE to TCE under the same conditions. Immobilization could be confirmed by using fluorescein-labeled antibody probes raised against D. multivorans. In granules obtained from the R1 reactor, D. multivorans grew mainly in microcolonies located in the centers of the granules, while in the test reactor, the bacterium mainly covered the surfaces of granules. PMID:9572963

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

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 ...

Application of safety checklist to the analysis of the IEA-R1 reactor water retreatment system; Utilizacao do checklist de seguranca na analise do sistema de retratamento de agua do reator IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Sauer, Maria Eugenia Lago Jacques; Sara Neto, Antonio Jorge; Lima, Toni Carlos Caboclo de; Ribeiro, Maria Alice Morato [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: melsauer@ipen.br

2005-07-01

In 1999, the management of the IEA-R1 Research Reactor (pool type - 5 MWth), located at IPEN/CNEN-SP, started the evaluation of the Reactor Pool Water Retreatment System to identify operational aspects, which could compromise the operators safety. The purpose was to identify and propose enhancements to the system which would be installed to substitute for the existing one. This process was conducted through a qualitative study of the system in operation. This study was carried out by a team composed of specialists in reactor operation, systems maintenance and radiological protection, and one safety analyst. The study consisted, basically, in local inspections to verify the physical and operational conditions of each equipment / component as well as aspects related to maintenance activities of the system. The process control and the operator procedures associated with the retreatment of the reactor pool water were also reviewed. The methodology adopted to develop the study was based in process hazard analysis technique named Safety Checklist. This paper presents a summary of this study and the main results obtained. Some operational and safety problems identified, the prevention and/or correction means to avoid them, and the recommendations and suggestions that have been implemented to the new design of the IEA-R1 Reactor Water Retreatment System, whose installation was concluded in 2003, are also presented. (author)

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

Characteristics of D-3He fueled frc reactor: ARTEMIS-L

International Nuclear Information System (INIS)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

International Nuclear Information System (INIS)

Hien, P.D.

1999-01-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

1999-08-01

The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

Development and implementation of a new pneumatic transfer system for materials irradiation at IEA-R1 reactor

International Nuclear Information System (INIS)

Fernando, Alberto de Jesus

2011-01-01

Pneumatic Transfer Systems (PTS) are classified as mechanical equipment largely operated all over the world for transport of a huge sort of objects, samples and materials located at nearly terminals or even at separated ones. System applicability is often recognized in many activities, such as medicine (hospital settings, clinical analysis labs), industry (steel, automobiles, mining, chemical, food, construction), trading (gas station, movies, supermarkets, banks, e-commerce) and federal agencies (post services, federal courts, public enterprises). In the nuclear settings, PTS shows also a vast array of applications, being a part of radioisotope production, as well as short-lived radiopharmaceuticals, including 67 Ga, 201 Tl, 18 F and 123 I-ultra pure. Besides, PTS are also used at radioactive waste management plants and research institutes that apply neutron activation analysis (NAA). This work was directed toward the design and operation of a new PTS for the IEA-R1 nuclear research reactor settled at Instituto de Pesquisas Energeticas e Nucleares (IPEN) for NAA application. With this aim, it was calculated the charge of reactor core grid plate and sample transport testing. Neutron flux at irradiating position was determined as 3,70 ± 0,26 10 12 n cm -2 s -1 . (author)

Obtaining of total and thermal neutron flux in the carousel facility of the TRIGA MARK IPR-R1 reactor using the Monte Carlo transport method

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2011-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The goal this work is modelling of the IPR-R1 Research Reactor TRIGA using the codes MCNPX2.6.0 (Monte Carlo N-Particle Transport extend) and MCNP5 to the calculating the neutron flux in the carousel facility. In each simulation the sample was placed in a different position, totaling forty positions around of the reactor core. The comparison between the results obtained with experimental values from other work showing a relatively good agreement. Moreover, this methodology is a theoretical tool in validating of the experimental values and necessary for determining neutron flux which can not be accessible experimentally. (author)

Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

International Nuclear Information System (INIS)

Frajndlich, R.; Sousa, J.A. de.

1985-01-01

A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author) [pt

Steviamine, a new class of indolizidine alkaloid [(1R,2S,3R,5R,8aR-3-hydroxymethyl-5-methyloctahydroindolizine-1,2-diol hydrobromide

Directory of Open Access Journals (Sweden)

Amber L. Thompson

2009-11-01

Full Text Available X-ray crystallographic analysis of the title hydrobromide salt, C10H20N+·Br−, of (1R,2S,3R,5R,8aR-3-hydroxymethyl-5-methyloctahydroindolizine-1,2-diol defines the absolute and relative stereochemistry at the five chiral centres in steviamine, a new class of polyhydroxylated indolizidine alkaloid isolated from Stevia rebaudiana (Asteraceae leaves. In the crystal structure, molecules are linked by intermolecular O—H...Br and N—H...Br hydrogen bonds, forming double chains around the twofold screw axes along the b-axis direction. Intramolecular O—H...O interactions occur.

Sources of gamma radiation in a reactor core

Energy Technology Data Exchange (ETDEWEB)

Roos, Matts

1959-05-15

In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

Factors affecting nuclear research reactor utilization across countries

International Nuclear Information System (INIS)

Hien, P.D.

2000-01-01

In view of the worldwide declining trend of research reactor utilization and the fact that many reactors in developing countries are under-utilised, a question naturally arises as to whether the investment in a research reactor is justifiable. Statistical analyses were applied to reveal relationships between the status of reactor utilization and socio-economic conditions among countries, that may provide a guidance for reactor planning and cost benefit assessment. The reactor power has significant regression relationships with size indicators such as GNP, electricity consumption and R and D expenditure. Concerning the effectiveness of investment in research reactors, the number of reactor operation days per year only weakly correlates with electricity consumption and R and D expenditure, implying that there are controlling factors specific of each group of countries. In the case of less developed countries, the low customer demands on reactor operation may be associated with the failure in achieving quality assurance for the reactor products and services, inadequate investment in the infrastructure for reactor exploitation, the shortage of R and D funding and well trained manpower and the lack of measures to get the scientific community involved in the application of nuclear techniques. (author)

Integral Fast Reactor Program

International Nuclear Information System (INIS)

Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

1993-06-01

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

International Nuclear Information System (INIS)

Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

2010-01-01

Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

The fusion reactor

International Nuclear Information System (INIS)

Brennan, M.H.

1974-01-01

Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

Annual report of Department of Research Reactor, 1997. April 1, 1997 - March 31, 1998

International Nuclear Information System (INIS)

1999-01-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of the department in fiscal year of 1997 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses

International Nuclear Information System (INIS)

Rios, Ilka Antonia

2013-01-01

This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

Mineralization and defluoridation of 2,2,3,3-tetrafluoro -1-propanol (TFP) by UV oxidation in a novel three-phase fluidized bed reactor (3P-FBR).

Science.gov (United States)

Shih, Yu-Jen; Tsai, Meng-Tso; Huang, Yao-Hui

2013-05-01

2,2,3,3-Tetrafluoro-1-propanol (TFP, C3H4F4O, M.W. = 132.06) is extensively used as the solvent in CD-R and DVD-R fabrication. Since it has a fluorinated alky-chain configuration and is non-biodegradable, its treatment by conventional oxidation methods is typically very inefficient. In this work, novel three-phase fluidized bed reactor (3P-FBR, 7.5 cm in diameter, 50 cm high) that combines photo oxidation (UV/H2O2, one of AOPs (Advanced Oxidation Process) and adsorption (BT5 iron oxide as adsorbent) processes is designed for mineralizing and defluorinizing TFP wastewater. The experimental results reveal that TFP can be efficiently mineralized, and the BT5 that is circulated by aeration in the 3P-FBR system can remove the released fluoride ions in the reaction period. Irradiation with 254 nm UV and a 10 mM H2O2 dose yield a TOC removal of TFP (1.39 mM, equivalent to an initial TOC of 50 ppm) of over 99.95% in 2 h, and 99% of fluoride was removed by BT5 with an adsorption capacity of 24.1 mg-F g(-1). Copyright © 2013 Elsevier Ltd. All rights reserved.

Development Plan and R&D Status of China Lead-based Reactors (CLEAR) for ADS, LFR and Fusion

International Nuclear Information System (INIS)

Wu Yican

2013-01-01

China has launched the ADS engineering construction project in 2011. The engineering design and related R&D activities are going on in order to finish the construction of the first system around 2017. China has a strong program to support the development of fusion and hybrid concepts and R&D activities in order to initiate the construction of fusion test reactor in the near future. CLEAR may play an important bridge role in the transition period from fission energy to fusion energy, such as to support: • Nuclear waste transmutation, fuel breeding, energy production, for promoting fission industry. • Technology sharing, pre-test platform, tritium supply, for promoting fusion development

Annual report of department of research reactors, 2001. April 1, 2001 - March 31, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-12-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2001 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

Removal of the Materials Test Reactor overhead working reservoir

International Nuclear Information System (INIS)

Lunis, B.C.

1975-10-01

Salient features of the removal of an excessed contaminated facility, the Materials Test Reactor (MTR) overhead working reservoir (OWR) from the Test Reactor Area to the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory are described. The 125-ton OWR was an overhead 160,000-gallon-capacity tank approximately 193 feet high which supplied cooling water to the MTR. Radiation at ground level beneath the tank was 5 mR/hr and approximately 600 mR/hr at the exterior surface of the tank. Sources ranging from 3 R/hr to in excess of 500 R/hr exist within the tank. The tank interior is contaminated with uranium, plutonium, and miscellaneous fission products. The OWR was lowered to ground level with the use of explosive cutters. Dismantling, decontamination, and disposal were performed by Aerojet Nuclear Company maintenance forces

Crystal structure of (1S,3R,8R,9R-2,2-dichloro-3,7,7-trimethyl-10-methylenetricyclo[6.4.0.01,3]dodecan-9-ol

Directory of Open Access Journals (Sweden)

Ahmed Benzalim

2016-08-01

Full Text Available The title compound, C16H24Cl2O, was synthesized by treating (1S,3R,8S,9R,10S-2,2-dichloro-3,7,7,10-tetramethyl-9,10-epoxytricyclo[6.4.0.01,3]dodecane with a concentrated solution of hydrobromic acid. It is built up from three fused rings: a cycloheptane ring, a cyclohexyl ring bearing alkene and hydroxy substituents, and a cyclopropane ring bearing two chlorine atoms. The asymmetric unit contains two molecules linked by an O—H...O hydrogen bond. In the crystal, further O—H...O hydrogen bonds build up an R44(8 cyclic tetramer. One of the molecules presents disorder that affects the seven-membered ring. In both molecules, the six-membered rings display a chair conformation, whereas the seven-membered rings display conformations intermediate between boat and twist-boat for the non-disordered molecule and either a chair or boat and twist-boat for the disordered molecule owing to the disorder. The absolute configuration for both molecules is 1S,3R,8R,9R and was deduced from the chemical pathway and further confirmed by the X-ray structural analysis.

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

Characterization of filters cartridges from the water polishing system of IEA-R1 reactor: radiometric methods

International Nuclear Information System (INIS)

Tessaro, Ana Paula G.; Vicente, Roberto

2015-01-01

The acceptance of radioactive waste in a repository depends primarily on knowledge of the radioisotopic inventory of the material, according to regulations established by regulatory agencies. The primary characterization is also a fundamental action to determine further steps in the management of the radioactive wastes. The aim of this work is to report the development of non-destructive methods for primary characterization of filters cartridges discarded as radioactive waste. The filters cartridges are used in the water polishing system of the IEA-R1 reactor retaining the particles in suspension in the reactor cooling water. The IEA-R1 is a pool type reactor with a thermal power of 5 MW, moderated and cooled with light water. It is located in the Energy and Nuclear Research Institute (IPEN-CNEN), in São Paulo, Brazil. The cartridge filters become radioactive waste when they are saturated and do not meet the required flow for the proper operation of the water polishing system. The activities of gamma emitters present in the filters are determined using gamma spectrometry, dose rate measurements and the Point Kernel Method to correlate results from both measurements. For the primary characterization, one alternative method is the radiochemical analysis of slices taken from each filter, what presents the disadvantage of higher exposures personnel and contamination risks. Another alternative method is the calibration of the measurement geometry of a gamma spectrometer, which requires the production of a standard filter. Both methods are necessary but can not be used in operational routine of radioactive waste management owing to cost and complexity. The method described can be used to determine routinely the radioactive inventory of these filters and other radioactive wastes, avoiding the necessity of destructive radiochemical analysis, or the necessity of calibrating the geometry of measurement. (author)

Status of EC solid breeder blanket designs and R and D for demo fusion reactors

International Nuclear Information System (INIS)

Proust, E.; Anzidei, L.; Moons, F.

1994-01-01

Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

Measurement of thermal, epithermal and fast neutron flux in the IEA-R1 reactor by the foil activation method

International Nuclear Information System (INIS)

Koskinas, M.F.

1979-01-01

Experimental and theoretical details of the foil activation method applied to neutrons flux measurements at the IEA-R1 reactor are presented. The thermal - and epithermal - neutron flux were determined form activation measurements of gold, cobalt and manganese foils; and for the fast neutron flux determination, aluminum, iron and nickel foils were used. The measurements of the activity induced in the metal foils were performed using a Ge-Li gamma spectrometry system. In each energy range of the reactor neutron spectrum, the agreement among the experimental flux values obtained using the three kind of materials, indicates the consistency of the theoretical approach and of the nuclear parameters selected. (Author) [pt

The neutron and gamma-ray dose characterization using the Monte Carlo method to study the feasibility of the Prompt Gamma Activation Analysis technique at IPR-R1 TRIGA reactor in Brazil

Energy Technology Data Exchange (ETDEWEB)

Guerra, Bruno T.; Soares, Alexandre L.; Grynberg, Suely E.; Menezes, Maria Angela B.C., E-mail: brunoteixeiraguerra@yahoo.com.br, E-mail: menezes@cdtn.br, E-mail: asleal@cdtn.br, E-mail: seg@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2013-07-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in {sup 235}U. The Implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) Technical at the TRIGA IPR-R1 research reactor of the CDTN will significantly increase in the types of matrices analyzable. A project is underway in order to implement this technique in CDTN. In order of verified the feasibility of the PGNAA at the TRIGA reactor, the MCNP (Monte Carlo N-Particle) method is used to theoretical calculations. This paper presents the results of a preliminary study of the neutron and gamma-ray dose in the room where the reactor is located, in case of implementation of this technique in the IPR-R1. (author)

The neutron and gamma-ray dose characterization using the Monte Carlo method to study the feasibility of the Prompt Gamma Activation Analysis technique at IPR-R1 TRIGA reactor in Brazil

International Nuclear Information System (INIS)

Guerra, Bruno T.; Soares, Alexandre L.; Grynberg, Suely E.; Menezes, Maria Angela B.C.

2013-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The Implementation of the PGNAA (Prompt Gamma Neutron Activation Analysis) Technical at the TRIGA IPR-R1 research reactor of the CDTN will significantly increase in the types of matrices analyzable. A project is underway in order to implement this technique in CDTN. In order of verified the feasibility of the PGNAA at the TRIGA reactor, the MCNP (Monte Carlo N-Particle) method is used to theoretical calculations. This paper presents the results of a preliminary study of the neutron and gamma-ray dose in the room where the reactor is located, in case of implementation of this technique in the IPR-R1. (author)

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

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

Design and R and D activities on ceramic breeder blanket for fusion experimental reactors in JAERI

International Nuclear Information System (INIS)

Kurasawa, T.; Takatsu, H.; Sato, S.; Nakahira, M.; Furuya, K.; Hashimoto, T.; Kawamura, H.; Kuroda, T.; Tsunematsu, T.; Seki, M.

1995-01-01

Design and R and D activities on ceramic breeder blanket of a fusion experimental reactor have been progressed in JAERI. A layered pebble bed type ceramic breeder blanket with water cooling is a prime candidate concept. Design activities have been concentrated on improvement of the design by conducting detailed analyses and also by fabrication procedure consideration based on the current technologies. A wide variety of R and Ds have also been conducted in accordance with the design activities. Development of fabrication technology of the blanket box structure and its mechanical testing, elementary testing on thermal performances of the pebble bed, and engineering-oriented material tests of breeder and beryllium pebbles are the main achievements during the last two years. (orig.)

Evaluation of the gait performance of above-knee amputees while walking with 3R20 and 3R15 knee joints

Directory of Open Access Journals (Sweden)

AliReza Taheri

2012-01-01

Full Text Available Background: The performance of the subjects with above-knee amputation is noticeably poorer than normal subjects. Various types of components have been designed to compensate their performance. Among various prosthetic components, the knee joint has great influence on the function. Two types of knee joints (3R15, 3R20 have been used broadly for above-knee prostheses. However, there is not enough research to highlight the influence of these joints on the gait performance of the subjects. Therefore, an aim of this research was to investigate the performance of the above-knee amputees while walking with 3R15 and 3R20 knee joints. Materials and Methods: 7 above-knee amputees were recruited in this research study. They were asked to walk with a comfortable speed to investigate the gait function of the subjects with 3 cameras 3D motion analysis system (Kinematrix system. The difference between the performances of the subjects with these joints was compared by use of paired t-test. Results: The results of this study showed that, the performances of the subjects with 3R20 were better than that with 3R15. The walking speed of the subjects with 3R20 was 66.7 m/min compared to 30.4 m/min (P-value = 0.045. Moreover; the symmetry of walking with 3R20 was more than that with 3R15, based on the spatio- temporal gait parameters values (P-value <0.05. Conclusion: The difference between the performances of the subjects with 3R20 and 3R15 knee joints was related to the walking speed, which improved while walking with 3R20 joint.

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Osiris reactor descriptive report

International Nuclear Information System (INIS)

1976-03-01

OSIRIS is a swimming pool reactor of 70 MW thermal power. Its main purpose is the irradiation of reactor materials in high neutron flux. A description is given of the air conditioning, ventilation, and radioactive gas removal system. (R.L.)

Evaluation of the physical protection system of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Vaz, Antonio C.A.; Conti, Thadeu das N.

2013-01-01

The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

Evaluation of the physical protection system of the IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Vaz, Antonio C.A.; Conti, Thadeu das N., E-mail: acavaz@ipen.br, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

R2/R0-WTR decommissioning cost. Comparison and benchmarking analysis

International Nuclear Information System (INIS)

Varley, Geoff; Rusch, Chris

2001-10-01

SKI charged NAC International with the task of determining whether or not the decommissioning cost estimates of R2/R0 (hereafter simply referred to as R2) and Aagesta research reactors are reasonable. The associated work was performed in two phases. The objective in Phase I was to make global comparisons of the R2 and Aagesta decommissioning estimates with the estimates/actual costs for the decommissioning of similar research reactors in other countries. This report presents the results of the Phase II investigations. Phase II focused on selected discrete work packages within the decommissioning program of the WTR reactor. To the extent possible a comparison of those tasks with estimates for the R2 reactor has been made, as a basis for providing an opinion on the reasonableness of the R2 estimate. The specific WTR packages include: reactor vessel and internals dismantling; biological shield dismantling; primary coolant piping dismantling; electrical equipment removal; waste packaging; transportation and disposal of radioactive concrete and reactor components; project management, licensing and engineering; and removal of ancillary facilities. The specific tasks were characterised and analysed in terms of fundamental parameters including: task definition; labour hours expended; labour cost; labour productivity; length of work week; working efficiency; working environment and impact on job execution; external costs (contract labour, materials and equipment); total cost; waste volumes; and waste packaging and transport costs. Based on such detailed raw data, normalised unit resources have been derived for selected parts of the decommissioning program, as a first step towards developing benchmarking data for D and D activities at research reactors. Several general conclusions emerged from the WTR decommissioning project. Site characterisation can confirm or negate major assumptions, quantify waste volumes, delineate obstacles to completing work, provide an understanding

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

The role of acid incubation in rapid immobilization of hydrogen-producing culture in anaerobic upflow column reactors

Energy Technology Data Exchange (ETDEWEB)

Zhang, Zhen-Peng; Tay, Joo-Hwa [School of Civil and Environmental Engineering, Nanyang Technological University (Singapore); Institute of Environmental Science and Engineering, Nanyang Technological University (Singapore); Show, Kuan-Yeow [Faculty of Science, Engineering and Technology, University Tunku Abdul Rahman, 31900 Kampar, Perak (Malaysia); Liang, David Tee [Institute of Environmental Science and Engineering, Nanyang Technological University (Singapore); Lee, Duu-Jong [Department of Chemical Engineering, National Taiwan University, Taipei 10617 (China); Su, Ay [Department of Mechanical Engineering, Fuel Cell Center, Yuan-Ze University, Taoyuan 320 (China)

2008-10-15

An approach of acidification was examined on formation of hydrogen-producing granules and biofilms in upflow column-shaped reactors. The reactors were fed with synthetic glucose wastewater and operated at 37 C and pH 5.5. The acclimated anaerobic culture was inoculated in four reactors designated R1, R2, R3 and R4, with R3 and R4 filled with granular activated carbon as support medium. To unveil the roles of acidification, microbial culture in R2 and R3 was subject to an acid incubation for 24 h by shifting the culture pH from 5.5 to 2.0. The experimental results suggested that the acidification substantially accelerated microbial granulation, but not biofilm formation. Microbial activities were inhibited by the acid incubation for about 78 h, resulting in the retarded formation of biofilms of the acidified culture. Reducing culture pH resulted in improvement in cell surface physicochemical properties favoring microbial adhesion and immobilization. Zeta potential increased from -25.3 mV to 11.9 mV, hydrophobicity in terms of contact angle improved from 31 to 38 and production of extracellular polymers increased from 66 mg/g-VSS to 136 mg/g-VSS. As a result of the formation of granules and biofilms, high hydrogen production rates of 6.98 and 7.49 L/L h were achieved in granule-based and biofilm-based reactors, respectively. It is concluded that acid incubation is an efficient means to initiate the rapid formation of granules by regulating the surface characteristics of microbial culture. The use of support media as starting nuclei may result in rapid formation of biofilms without the acidification. (author)

The role of acid incubation in rapid immobilization of hydrogen-producing culture in anaerobic upflow column reactors

International Nuclear Information System (INIS)

Zhang, Zhen-Peng; Tay, Joo-Hwa; Show, Kuan-Yeow; Liang, David Tee; Lee, Duu-Jong; Su, Ay

2008-01-01

An approach of acidification was examined on formation of hydrogen-producing granules and biofilms in upflow column-shaped reactors. The reactors were fed with synthetic glucose wastewater and operated at 37 C and pH 5.5. The acclimated anaerobic culture was inoculated in four reactors designated R1, R2, R3 and R4, with R3 and R4 filled with granular activated carbon as support medium. To unveil the roles of acidification, microbial culture in R2 and R3 was subject to an acid incubation for 24 h by shifting the culture pH from 5.5 to 2.0. The experimental results suggested that the acidification substantially accelerated microbial granulation, but not biofilm formation. Microbial activities were inhibited by the acid incubation for about 78 h, resulting in the retarded formation of biofilms of the acidified culture. Reducing culture pH resulted in improvement in cell surface physicochemical properties favoring microbial adhesion and immobilization. Zeta potential increased from -25.3 mV to 11.9 mV, hydrophobicity in terms of contact angle improved from 31 to 38 and production of extracellular polymers increased from 66 mg/g-VSS to 136 mg/g-VSS. As a result of the formation of granules and biofilms, high hydrogen production rates of 6.98 and 7.49 L/L h were achieved in granule-based and biofilm-based reactors, respectively. It is concluded that acid incubation is an efficient means to initiate the rapid formation of granules by regulating the surface characteristics of microbial culture. The use of support media as starting nuclei may result in rapid formation of biofilms without the acidification. (author)

Remote level radiation monitoring system for the brazilian IEA-R1 nuclear research reactor for routine radiation protection procedures and as a support tool in case of radiological emergency

International Nuclear Information System (INIS)

Cardenas, Jose P.N.; Romero Filho, Christovam R.; Madi Filho, Tufic

2008-01-01

Nuclear facilities must monitoring radiation levels to establish procedures for radiological protection staff involving workers and the public. The Instituto de Pesquisas Energeticas e Nucleares - IPEN has 5 important plants and in case of accident in one of them, the Institute keeps operational an Emergency Response Plan (ERP). This document (ERP) is designed to coordinate all procedures to assure safe and secure conditions for workers, environment and the public. One of this plants is the IEA-R1 reactor, it is the oldest nuclear research reactor (pool type) in Latin America, reached it first criticality in September of 1957. The reactor is used 60 hours/week with continuous operation and with nominal power of 3.5 MW, with technical conditions to operate at 5 MW thermal power. This reactor has a Radiological Emergency Plan that establishes the implementation of rules for workers and people living at the exclusion area in the case of an emergency situation. This paper aims to describe the implementation of a computational system developed for remote radiation monitoring, in a continuous schedule of IEA-R1 nuclear research reactor containment building. Results of this action can be used as a support mean in a radiological emergency. All necessary modules for radiation detection, signals conditioners and processing, data acquisition board, software development and computer specifications are described. The data acquisition system operating in the reactor shows readings concerned to radiation environment such as activity, doses and concentration in real time and displays a periodical data bank (Data Base) of this features allowing through the surveillance of the operation records anytime, leading to studies and analysis of radiation levels. Results of this data acquisition are shown by means of computer graphics screens developed for windows environment using Visual Basic software. (author)

Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

2005-01-01

The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

Energy Technology Data Exchange (ETDEWEB)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

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

International Nuclear Information System (INIS)

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

1982-03-01

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

Open-ended fusion devices and reactors

International Nuclear Information System (INIS)

Kawabe, T.; Nariai, H.

1983-01-01

Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Annual report of Department of Research Reactor, 2003. April 1, 2003 - March 31, 2004

International Nuclear Information System (INIS)

2005-02-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides the RI production and its R and D are carried out. This report describes the activities of the department in fiscal year of 2003 and also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, utilization of irradiation and neutron beam experiments, technical management including management of fuels and water chemistry, and related R and D works. The RI production and its R and D works were conducted as well. The international co-operations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

A study on future nuclear reactor technology and development strategy

Energy Technology Data Exchange (ETDEWEB)

Kim, S. Y.; Kim, S. H.; Sohn, D. S.; Suk, S. D.; Zee, S. K.; Yang, M. H.; Kim, H. J.; Park, W. S

2000-12-01

Development of nuclear reactor and fuel cycle technology for future is essential to meet the current issues such as enhancement of nuclear power reactor safety, economically competitive with gas turbine power generation, less production of radioactive waste, proliferation resistant fuel cycle, and public acceptance in consideration of lack of energy resources in the nuclear countries worldwide as well as in Korea. This report deals with as follows, 1) Review the world energy demand and supply perspective and analyse nature of energy and sustainable development to set-up nuclear policy in Korea 2) Recaptitulate the current long term nuclear R and D activities 3) Review nuclear R and D activities and programs of USA, Japan, France, Russia, international organizations such as IAEA, OECD/NEA 4) Recommend development directions of nuclear reactors and fuels.

A study on future nuclear reactor technology and development strategy

International Nuclear Information System (INIS)

Kim, S. Y.; Kim, S. H.; Sohn, D. S.; Suk, S. D.; Zee, S. K.; Yang, M. H.; Kim, H. J.; Park, W. S.

2000-12-01

Development of nuclear reactor and fuel cycle technology for future is essential to meet the current issues such as enhancement of nuclear power reactor safety, economically competitive with gas turbine power generation, less production of radioactive waste, proliferation resistant fuel cycle, and public acceptance in consideration of lack of energy resources in the nuclear countries worldwide as well as in Korea. This report deals with as follows, 1) Review the world energy demand and supply perspective and analyse nature of energy and sustainable development to set-up nuclear policy in Korea 2) Recaptitulate the current long term nuclear R and D activities 3) Review nuclear R and D activities and programs of USA, Japan, France, Russia, international organizations such as IAEA, OECD/NEA 4) Recommend development directions of nuclear reactors and fuels

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

Antiausterity activity of arctigenin enantiomers: importance of (2R,3R)-absolute configuration.

Science.gov (United States)

Awale, Suresh; Kato, Mamoru; Dibwe, Dya Fita; Li, Feng; Miyoshi, Chika; Esumi, Hiroyasu; Kadota, Shigetoshi; Tezuka, Yasuhiro

2014-01-01

From a MeOH extract of powdered roots of Wikstroemia indica, six dibenzyl-gamma-butyrolactone-type lignans with (2S,3S)-absolute configuration [(+)-arctigenin (1), (+)-matairesinol (2), (+)-trachelogenin (3), (+)-nortrachelogenin (4), (+)-hinokinin (5), and (+)-kusunokinin (6)] were isolated, whereas three dibenzyl-gamma-butyrolactone-type lignans with (2R,3R)-absolute configuration [(-)-arctigenin (1*), (-)-matairesinol (2*), (-)-trachelogenin (3*)] were isolated from Trachelospermum asiaticum. The in vitro preferential cytotoxic activity of the nine compounds was evaluated against human pancreatic PANC-1 cancer cells in nutrient-deprived medium (NDM), but none of the six lignans (1-6) with (2S,3S)-absolute configuration showed preferential cytotoxicity. On the other hand, three lignans (1*-3*) with (2R,3R)-absolute configuration exhibited preferential cytotoxicity in a concentration-dependent manner with PC50 values of 0.54, 6.82, and 5.85 microM, respectively. Furthermore, the effect of (-)- and (+)-arctigenin was evaluated against the activation of Akt, which is a key process in the tolerance to nutrition starvation. Interestingly, only (-)-arctigenin (1*) strongly suppressed the activation of Akt. These results indicate that the (2R,3R)-absolute configuration of (-)-enantiomers should be required for the preferential cytotoxicity through the inhibition of Akt activation.

Studsvik`s fuel R and D projects

Energy Technology Data Exchange (ETDEWEB)

Grounes, M [Studsvik Nuclearr AB, Nykoping (Sweden)

1997-08-01

The report reviews some recently performed, ongoing and planned fuel R and D projects, executed by Studsvik Nuclear AB, a subsidiary of Studsvik AB. Data from these projects are used as experimental support for fuel modelling at high burnup. Much of Studsvik Nuclear`s R and D work has been concentrated on fuel testing, which can be made in the R2 test reactor with high precision under realistic water reactor conditions. This type of work started in the early 1960s. The fuel testing projects executed at Studsvik have been organized under three different types of sponsorship: International (multilateral) fuel projects: jointly sponsored internationally on a world-wide basis, with project information remaining restricted to the project participants throughout the project`s duration and for some pre-determined time after project completion; Bilateral fuel projects: sponsored by one single organization, or a few co-operating organizations, with project information remaining restricted to the sponsor, sometimes published later; in-house R and D work: sponsored by Studsvik Nuclear. The fuel testing activities can be divided into a number of well-defined steps as follows: Base irradiation, performed in a power reactor, or in Studsvik`s R2 test reactor; power ramping and/or other in-pile measurements, performed in Studsvik`s R2 test reactor. Non-destructive testing between different phases of an experiment, performed in Studsvik`s R2 reactor pool, or in Studsvik`s Hot Cell Laboratory; destructive post-irradiation examinations, performed in Studsvik`s Hot Cell Laboratory, or in the sponsor`s hot cell laboratory. 47 refs, 2 tabs.

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.)

Anticoking Coatings for High Temperature Petrochemical Reactors Revêtements pour réacteurs pétrochimiques à cokage réduit

Directory of Open Access Journals (Sweden)

Ropital F.

2006-12-01

Full Text Available Coke deposition is a major problem for several refinery and petrochemical processes, among which steam cracking for ethylene production. Decreasing coking rates will reduce decoking frequency and will also decrease heat transfer degradation. For these processes, a means of reducing coking phenomena is the application of anticoking coatings on the reactor walls. For this purpose a methodology was developed under a CEC Brite-Euram project with steam cracking as the main application. A first selection of the coatings and their techniques of deposition was performed according to their composition (non catalytic compounds, their resistance to carburization and oxidation and also according to the nature of the substrate. The feasibility of coating deposition was studied for small diameter furnace tubes and the techniques were perfected in particular with respect to the physicochemical characterization of the coatings and their resistance during thermal cycling and coking-decoking cycles. For the thermal cycling and the coking evaluation, industrial conditions were reproduced as closely as possible. For coking tests, a microreactor with complete mixing of the gas phase was developed for temperatures up to 950°C. The coking behavior of the coated samples was compared with conventional refractory alloys: reduction of the coking rate by a factor of 3 was measured with the most promising coating. This work was completed by the kinetic modelling of coking on uncoated materials. La formation de coke est un problème crucial pour de nombreux procédés pétrochimiques. Ralentir le cokage permet de réduire la fréquence des décokages et de limiter la dégradation du transfert thermique. Un des moyens envisageables pour réduire le cokage est le dépôt d'un revêtement à la paroi de ces réacteurs. La mise au point de tels revêtements a fait l'objet d'un projet de recherche européen Brite-Euram, pour lequel le vapocraquage a été la principale

Simultaneous Coproduction of Hydrogen and Ethanol in Anaerobic Packed-Bed Reactors

Directory of Open Access Journals (Sweden)

Cristiane Marques dos Reis

2014-01-01

Full Text Available This study evaluated the use of an anaerobic packed-bed reactor for hydrogen production at different hydraulic retention times (HRT (1–8 h. Two reactors filled with expanded clay and fed with glucose (3136–3875 mg L−1 were operated at different total upflow velocities: 0.30 cm s−1 (R030 and 0.60 cm s−1 (R060. The effluent pH of the reactors was maintained between 4 and 5 by adding NaHCO3 and HCl solutions. It was observed a maximum hydrogen production rate of 0.92 L H2 h−1 L−1 in R030 at HRT of 1 h. Furthermore, the highest hydrogen yield of 2.39 mol H2 mol−1 glucose was obtained in R060. No clear trend was observed by doubling the upflow velocities at this experiment. High ethanol production was also observed, indicating that the ethanol-pathway prevailed throughout the experiment.

Nondestructive inspection of the tubes of TRIGA IPR-R1 reactor heat exchanger by eddy current testing

International Nuclear Information System (INIS)

Silva Junior, Silverio F.; Silva, Roger F.; Oliveira, Paulo F.; Barreto, Erika S.; Ribeiro, Isabela G.; Fraiz, Felipe C.

2013-01-01

The IPR-R1 TRIGA MARK 1 reactor is an open pool type reactor, cooled light water. It is used for research activities, personnel training and radioisotopes production, in operation since 1960 at the Nuclear Technology Development Center - CDTN/CNEN. It operates at a maximum thermal power of 100 kW and usually, the fuel cooling is done by natural circulation. If necessary, an external auxiliary cooling system, with a shell-and-tube type heat exchanger, can be used to improve the water heat removal. As part of the ageing management program of the reactor, a nondestructive evaluation of their heat exchanger stainless steel tubes will be performed, in order to verify its integrity. The examinations will be performed using the eddy current test method, which allows the detection and characterization of structural discontinuities in the wall of the tubes, if existing. For this purpose, probes and reference standards were designed and manufactured at CDTN facilities and test procedures were established and validated. In this paper, a description of the proposed infrastructure as well as the test methodology to be used in the examinations are presented and discussed. (author)

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

Kinetics of the R + HBr {r_reversible} RH + Br (CH{sub 3}CHBr, CHBr{sub 2} or CDBr{sub 2}) equilibrium. Thermochemistry of the CH{sub 3}CHBr and CHBr{sub 2} radicals

Energy Technology Data Exchange (ETDEWEB)

Seetula, Jorma A. [Laboratory of Physical Chemistry, P.O. Box 55 (A.I. Virtasen aukio 1), FIN-00014 University of Helsinki, Helsinki (Finland)], E-mail: j.seetula@kolumbus.fi; Eskola, Arkke J. [Laboratory of Physical Chemistry, P.O. Box 55 (A.I. Virtasen aukio 1), FIN-00014 University of Helsinki, Helsinki (Finland)

2008-07-03

The kinetics of the reaction of the CH{sub 3}CHBr, CHBr{sub 2} or CDBr{sub 2} radicals, R, with HBr have been investigated in a temperature-controlled tubular reactor coupled to a photoionization mass spectrometer. The CH{sub 3}CHBr (or CHBr{sub 2} or CDBr{sub 2}) radical was produced homogeneously in the reactor by a pulsed 248 nm exciplex laser photolysis of CH{sub 3}CHBr{sub 2} (or CHBr{sub 3} or CDBr{sub 3}). The decay of R was monitored as a function of HBr concentration under pseudo-first-order conditions to determine the rate constants as a function of temperature. The reactions were studied separately from 253 to 344 K (CH{sub 3}CHBr + HBr) and from 288 to 477 K (CHBr{sub 2} + HBr) and in these temperature ranges the rate constants determined were fitted to an Arrhenius expression (error limits stated are 1{sigma} + Student's t values, units in cm{sup 3} molecule{sup -1} s{sup -1}, no error limits for the third reaction): k(CH{sub 3}CHBr + HBr) = (1.7 {+-} 1.2) x 10{sup -13} exp[+ (5.1 {+-} 1.9) kJ mol{sup -1}/RT], k(CHBr{sub 2} + HBr) = (2.5 {+-} 1.2) x 10{sup -13} exp[-(4.04 {+-} 1.14) kJ mol{sup -1}/RT] and k(CDBr{sub 2} + HBr) = 1.6 x 10{sup -13} exp(-2.1 kJ mol{sup -1}/RT). The energy barriers of the reverse reactions were taken from the literature. The enthalpy of formation values of the CH{sub 3}CHBr and CHBr{sub 2} radicals and an experimental entropy value at 298 K for the CH{sub 3}CHBr radical were obtained using a second-law method. The result for the entropy value for the CH{sub 3}CHBr radical is 305 {+-} 9 J K{sup -1} mol{sup -1}. The results for the enthalpy of formation values at 298 K are (in kJ mol{sup -1}): 133.4 {+-} 3.4 (CH{sub 3}CHBr) and 199.1 {+-} 2.7 (CHBr{sub 2}), and for {alpha}-C-H bond dissociation energies of analogous compounds are (in kJ mol{sup -1}): 415.0 {+-} 2.7 (CH{sub 3}CH{sub 2}Br) and 412.6 {+-} 2.7 (CH{sub 2}Br{sub 2}), respectively.

Measurement of thermal, epithermal and fast neutrons fluxes by the activation foil method at IEA-R1 reactor

International Nuclear Information System (INIS)

Dias, M.S.; Koskinas, M.F.; Berretta, J.R.; Fratin, L.; Botelho, S.

1990-01-01

The thermal, epithermal and fast neutron fluxes have been determined experimentally by the activation foil method at position GI, located near the IEA-R1 reactor core. The reactions used were 197 Au (n,gamma) 198 Au, for thermal and epithermal neutrons and 27 Na (n,alpha) 24 Na, for fast neutrons. The activities were measured by the 4π(PC)β-γ coincidence method. (author)

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

Neutron activation analysis at CDTN/CNEN using the IPR-R1 Triga Mark I reactor

International Nuclear Information System (INIS)

Menezes, Maria Angela de B.C.; Maretti Junior, Fausto; Kastner, Geraldo Frederico; Amaral, Angela Maria; Souza, Wagner de

2009-01-01

This paper describes in summary the activities developed by the Laboratory for Neutron Activation Analysis since the starting up of the IPR-R1 TRIGA Mark I research reactor in 1960. This Laboratory is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for significant percentage of CDTN's analytical demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays the neutron activation analysis is responsible for 70% of the analytical demand and the k 0 - Instrumental method for 80% of this demand answering clients' request and researches. In Brazil, CDTN is the only Institute that fully masters the Instrumental Neutron Activation Analysis k 0 -method using its own nuclear reactor. (author)

Development of a computational program to planning and control of the IEA-R1 reactor maintenance

International Nuclear Information System (INIS)

Martins, Mauro Onofre; Madi Filho, Tufic

2013-01-01

Maintenance is an essential activity in nuclear reactors. The components of safety systems of an industrial plant should have a low probability of failure, especially if there is a high risk of accidents that may cause environmental damage. In nuclear facilities, the presence of security systems is a technical specification and a requirement for their license and operation. In order to manage the entire information flow from the maintenance of the IEA-R1, a computational program (software) was developed, which not only plans and control all the maintenance, but also updates the documents and records to safeguard the quality, ensuring the safe operation of the reactor. The software has access levels and provides detailed reports of all maintenance planned and implemented, together with an individual history of the equipment during its lifetime in the facility. This work presents all the stages of the software development, description, compatibility, application, advantages and results obtained experimentally. (author)

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Monitoring device for the stability of a reactor core

International Nuclear Information System (INIS)

Sakurai, Mikio; Yamauchi, Koki.

1983-01-01

Purpose: To avoid unnecessary limitation on the operation conditions for maintaining the reactor stability. Constitution: The reactor stability is judged by taking notice of the axial power distribution of the reactor and monitoring the same online. Specifically, signals are received from a plurality of local power distribution detectors arranged axially in the reactor core to calculate the axial power distribution in computer. Further, a certain distance L is set from the lower end of the reactor core and the total value S1 for the power distribution in the region below the set value L and the total value S2 for the region above the set value L are determined based on the thus calculated power distribution, to thereby determine the ratio: R = S1/S2 between them. Separately, a certain value r is previously determined based on analysis or experiment such as the result of operation. Then, R and r are compared in a comparator and an alarm is generated, if R >r, with respect to the stability. Since monitoring is made based on the actual index, the applicable range of the operation region can be extended. (Ikeda, J.)

AP1000R pressurised water reactor project in china advances toward completion

International Nuclear Information System (INIS)

Harrop, G.

2014-01-01

The AP1000 R pressurised water reactor (PWR) project in China is the first deployment of its first-of-a-kind Generation III+ technology, making it one of most internationally important and industry-significant new build projects. The innovative AP1000 PWR design contains advanced passive safety and performance features that involve fewer active safety components than a traditional plant, thereby reducing the site footprint. The AP1000 reactor is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission, and it has received an Interim Design Acceptance Confirmation from the Office for Nuclear Regulation and an Interim Statement of Design Acceptability from the Environment Agency in the United Kingdom. Construction and testing of dual AP1000 PWR units is currently in progress in each of two coastal sites in the People's Republic of China: Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Since the initial contract award in 2007, the Westinghouse Consortium has worked in concert with the owners to construct the plants using innovative structural and mechanical modules. Uniquely designed plant components and essential instrumentation and control systems have been manufactured, delivered, and installed at the plants. Numerous personnel, including future reactor operators, have been trained at both the Sanmen and Haiyang sites, and technology transfer of technical documents and computer codes is well underway. The commercial operation dates are now nearing for Sanmen Unit 1 and Haiyang Unit 1, the first two units scheduled for completion. Consequently, these units are now in advanced stages of completion and present activities include planning and preparation for pre-operational testing, system turnover, and commissioning leading to fuel load, and eventual commercial operation. These activities are pioneering, in that they have never before been performed for a new build of

Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

International Nuclear Information System (INIS)

Shiozawa, Shusaku

2003-01-01

On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

Review of the United Kingdom fast reactor programme - March 1986

International Nuclear Information System (INIS)

Bramman, J.I.; John, C.T.; Wheeler, R.C.

1986-01-01

The UK programme in the field of fast reactors has continued successfully towards the following main objectives, details of which are contained in subsequent sections of this report: (2) progress with the prototype fast reactor (PFR) which achieved its design power on 4 March 1985; (3) nuclear fuel reprocessing; (4) commercial design studies; (5) structural integrity of LMFBR during its lifetime; (6) R and D work on components of LMFBR; (7) materials study; (8) sodium chemistry; (9) reactor core and fuel design philosophy; (10) safety problems; (11) plant performance studies

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Theoretical studies aiming at the IEA-R1 reactor core conversion from high U-235 enrichment to low U-235 enrichment

International Nuclear Information System (INIS)

Frajndlich, R.

1982-01-01

The research reactors, of which the fuel elements are of MTR type, functions presently, almost in their majority with high U-235 enrichment. The fear that those fuel elements might generate a considerabLe proliferation of nuclear weapons rendered almost mandatory the conversion of highly enriched fuel elements to a low U-235 enrichment. As the IEA-R1 reactor of IPEN is operating with highly enriched fuel elements a study aiming at this conversion was done. The problems related to the conversion and the results obtained, demonstrated the technical viabilty for its realization. (E.G.) [pt

The structures of T6, T3R3 and R6 bovine insulin: combining X-ray diffraction and absorption spectroscopy

DEFF Research Database (Denmark)

Frankær, Christian Grundahl; Knudsen, Marianne Vad; Noren, Katarina

2012-01-01

The crystal structures of three conformations, T6, T3R3 and R6, of bovine insulin were solved at 1.40, 1.30 and 1.80 Å resolution, respectively. All conformations crystallized in space group R3. In contrast to the T6 and T3R3 structures, different conformations of the N-terminal B-chain residue Phe......B1 were observed in the R6 insulin structure, resulting in an eightfold doubling of the unit-cell volume upon cooling. The zinc coordination in each conformation was studied by X-ray absorption spectroscopy (XAS), including both EXAFS and XANES. Zinc adopts a tetrahedral coordination in all R3 sites...... molecules, as well as in other high-resolution insulin structures. As the radiation dose for XRD experiments is two orders of magnitude higher compared with that of XAS experiments, the single crystals were exposed to a higher degree of radiation damage that affected the zinc coordination in the T3 sites...