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Sample records for iii reactor frn

  1. Operation experience at the TRIGA Mark III reactor FRN within the last two years

    International Nuclear Information System (INIS)

    Rau, G.

    1976-01-01

    A survey is given of the history of FRN-reactor operation since the last TRIGA conference. In 1975, the reactor had been operated for 830 MWh and had been pulsed 1050 times. In the first half of 1976, an appreciable Increase of reactor activities took place. In this time interval, the integrated power amounted to 996 MWh and the number of released pulses to 1024. In 1975, an important event was the exchange of the defective rotary specimen rack against a water-operated type. For this purpose, the whole core had to be unloaded and the pool water had to be removed. In this connection, damages at the beam-port weldings had been discovered and had to be repaired immediately. In December 1975, a new core configuration had been set up by adding ten spare fuel elements into the F-resp. G-ring. The gain of excess reactivity amounted to approximately 1,8 $. This however allowed reactor operation without disturbance by Xe-poisoning only for a limited time interval. (author)

  2. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

    Demmeler, M.; Rau, G.; Strube, D.

    1982-01-01

    The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

  3. Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

    1974-07-01

    Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

  4. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

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

  5. Mixed field dosimetry in the FRN exposure-room

    International Nuclear Information System (INIS)

    Jentzsch, U.

    1976-01-01

    The twin ionization chamber technique, chemical dosimeter systems and silicon diodes have been used to determine the neutron and gamma components of the reactor field in the FRN exposure-room. The knowledge of these components are of fundamental importance for irradiation experiments in the biological research program started at the FRN. The chemical systems and silicon diodes have the advantage of low cost and simple handling. Furthermore, silicon diodes have negligible gamma response and the shift of neutron response with variation of neutron energy spectrum is small. The results of measurements with the different methods have been compared and discussed. (author)

  6. New irradiation devices at the FRN reactor

    International Nuclear Information System (INIS)

    Stark, W.

    1980-01-01

    In order to fulfill the experimental demands three additional devices were constructed and installed. The first is a vertical irradiation tube in air surrounded by a lead cylinder (in the irradiation position). The second device is a rabbit system ending within the graphite moderator of the thermal column. The third device is so called rotating disk assembly, built to replace the rotary specimen rack

  7. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  8. Reactor oscillator - I - III, Part III - Electronic device

    International Nuclear Information System (INIS)

    Lolic, B.; Jovanovic, S.

    1961-12-01

    This report describes functioning of the reactor oscillator electronic system. Two methods of oscillator operation were discussed. The first method is so called method of amplitude modulation of the reactor power, and the second newer method is phase method. Both methods are planned for the present reactor oscillator

  9. Reactor oscillator - I - III, Part I

    International Nuclear Information System (INIS)

    Lolic, B.

    1961-12-01

    Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction

  10. Recent operational history of the new Sandia Pulsed Reactor III (SPR III)

    International Nuclear Information System (INIS)

    Schmidt, T.R.; Estes, B.F.; Reuscher, J.A.

    1977-01-01

    The Sandia Pulsed Reactor III (SPR III) is a fast-pulse research reactor which was designed and built at Sandia Laboratories and achieved criticality in August 1975. The reactor is now characterized and is in an operational configuration. The core consists of 18 fuel plates (258 kg fuel mass) of fully enriched uranium alloyed with 10 wt.% molybdenum. It is arranged in an annular configuration with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a height of 35.9 cm. The reactor core uses reflectors of copper and aluminum for control and an external bolting arrangement to secure the fuel plates. SPR III and SPR II are operated on an interchangeable basis using the same facility and control system. As of June 1977, SPR III has had over 240 operations with core temperatures up to 541 0 C

  11. Gen-III/III+ reactors. Solving the future energy supply shortfall. The SWR-1000 option

    International Nuclear Information System (INIS)

    Stosic, Z.V.

    2006-01-01

    Deficiency of non-renewable energy sources, growing demand for electricity and primary energy, increase in population, raised concentration of greenhouse gases in the atmosphere and global warming are the facts which make nuclear energy currently the most realistic option to replace fossil fuels and satisfy global demand. The nuclear power industry has been developing and improving reactor technology for almost five decades and is now ready for the next generation of reactors which should solve the future energy supply shortfall. The advanced Gen-III/III+ (Generation III and/or III+) reactor designs incorporate passive or inherent safety features which require no active controls or operational intervention to manage accidents in the event of system malfunction. The passive safety equipment functions according to basic laws of physics such as gravity and natural convection and is automatically initiated. By combining these passive systems with proven active safety systems, the advanced reactors can be considered to be amongst the safest equipment ever made. Since the beginning of the 90's AREVA NP has been intensively engaged in the design of two advanced Gen-III+ reactors: (i) PWR (Pressurized Water Reactor) EPR (Evolutionary Power Reactor) and (ii) BWR (Boiling Water Reactor) SWR-1000. The SWR-1000 reactor design marks a new era in the successful tradition of BWR technology. It meets the highest safety standards, including control of a core melt accident. This is achieved by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation. A short construction period, flexible fuel cycle lengths and a high fuel discharge burn-up contribute towards meeting economic goals. The SWR-1000 completely fulfils international nuclear regulatory requirements. (author)

  12. Thermal spectra of the TRIGA Mark III reactor; El espectro termico del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Macias B, L.R.; Palacios G, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

  13. CANDU technology for generation III + AND IV reactors

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    2005-01-01

    Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

  14. SMORN-III benchmark test on reactor noise analysis methods

    International Nuclear Information System (INIS)

    Shinohara, Yoshikuni; Hirota, Jitsuya

    1984-02-01

    A computational benchmark test was performed in conjunction with the Third Specialists Meeting on Reactor Noise (SMORN-III) which was held in Tokyo, Japan in October 1981. This report summarizes the results of the test as well as the works made for preparation of the test. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  16. Reactor oscillator - I - III, Part III - Electronic device; Reaktorski oscilator - I-III, III Deo - Elektronski uredjaj

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report describes functioning of the reactor oscillator electronic system. Two methods of oscillator operation were discussed. The first method is so called method of amplitude modulation of the reactor power, and the second newer method is phase method. Both methods are planned for the present reactor oscillator.

  17. The ARIES-III D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1992-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

  18. RB research nuclear reactor, Annual report for 1984, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

    1984-01-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

  19. RB research nuclear reactor - Annual report for 1986, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1987-01-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

  20. RB research nuclear reactor, Annual report for 1983, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

    1983-01-01

    The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

  1. RB research nuclear reactor, Annual report for 1989, I - III

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1989-12-01

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

  2. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  3. Thermal spectra of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Macias B, L.R.; Palacios G, J.

    1998-01-01

    The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

  4. Reactor oscillator - I - III, Part I; Reaktorski oscilator - I-III, I Deo

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction.

  5. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Energy Technology Data Exchange (ETDEWEB)

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  6. Validation of the Serpent 2-DYNSUB code sequence using the Special Power Excursion Reactor Test III (SPERT III)

    International Nuclear Information System (INIS)

    Knebel, Miriam; Mercatali, Luigi; Sanchez, Victor; Stieglitz, Robert; Macian-Juan, Rafael

    2016-01-01

    Highlights: • Full few-group cross section tables created by Monte Carlo lattice code Serpent 2. • Serpent 2 group constant methodology verified for HFP static and transient cases. • Serpent 2-DYNSUB tool chainvalidated using SPERT III REA experiments. • Serpent 2-DYNSUB tool chain suitable to model RIAs in PWRs. - Abstract: The Special Power Excursion Reactor Test III (SPERT III) is studied using the Serpent 2-DYNSUB code sequence in order to validate it for modeling reactivity insertion accidents (RIA) in PWRs. The SPERT III E-core was a thermal research reactor constructed to analyze reactor dynamics. Its configuration resembles a commercial PWR on terms of fuel type, choice of moderator, coolant flow and system pressure. The initial conditions of the rod ejection accident experiments (REA) performed cover cold startup, hot startup, hot standby and operating power scenarios. Eight of these experiments were analyzed in detail. Firstly, multi-dimensional nodal diffusion cross section tables were created for the three-dimensional reactor simulator DYNSUB employing the Monte Carlo neutron transport code Serpent 2. In a second step, DYNSUB stationary simulations were compared to Monte Carlo reference three-dimensional full scale solutions obtained with Serpent 2 (cold startup conditions) and Serpent 2/SUBCHANFLOW (operating power conditions) with a good agreement being observed. The latter tool is an internal coupling of Serpent 2 and the sub-channel thermal-hydraulics code SUBCHANFLOW. Finally, DYNSUB was utilized to study the eight selected transient experiments. Results were found to match measurements well. As the selected experiments cover much of the possible transient (delayed super-critical, prompt super-critical and super-prompt critical excursion) and initial conditions (cold and hot as well as zero, little and full power reactor states) one expects in commercial PWRs, the obtained results give confidence that the Serpent 2-DYNSUB tool chain is

  7. Preparations for decommissioning the TRIGA Mark III Berkeley Research Reactor

    International Nuclear Information System (INIS)

    Denton, Michael M.; Lim, Tek. H.

    1988-01-01

    On December 20, 1986 the chancellor of UC Berkeley announced his decision to decommission the 20 year old Berkeley Research Reactor citing as principal reasons a decline in use and a need to erect a new computer science building over the reactor's site. In order to meet the University's construction timetable for the new building, the reactor staff together with other units of the campus administration have initiated a program to remove the reactor structure and clear the room for unlicensed use as expediently as possible. Due to the sequence of events which must occur in a limited amount of time, the University adopted a policy to contract out as much of the work as possible, including generation of the defueling and decommissioning plans.The first physical step in the decommissioning project is the removal of the irradiated fuel. This task is largely contracted out to a commercial firm with experience in the transport of radioactive materials and reactor fuel. As suggested by the NRC, the reactor will be defueled under the current operating license. This requires that all fuel must be off-site before the DP can be approved. Therefore any delay in defueling in-turn delays the decommissioning. The NRC has given no commitment or date for completion of their review. Informal discussion with NRC project managers and the experience from other facilities indicate that the review process will take between six and nine months

  8. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR

    International Nuclear Information System (INIS)

    Franco C, A.

    1997-01-01

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author)

  9. Design and initial performance of the Sandia Pulsed Reactor-III

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Estes, B.F.

    1976-01-01

    The Sandia Pulsed Reactor-III (SPR-III) is a new fast pulsed reactor which has recently undergone initial testing at Sandia Laboratories. SPR-III is a uranium-10 weight percent molybdenum fuel assembly with a 17.78 cm irradiation cavity similar in design to SPR-II which has been in operation since 1967. The basic SPR-III design utilizes the same split-core configuration which has been proven with SPR-II; however, SPR-III uses external reflectors for control and external bolts to hold the fuel plates together. The core consists of sixteen fuel plates with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a core height of 31.9 cm. The fuel mass is about 227 kg of fully enriched uranium-10 weight percent molybdenum alloy. SPR III has completed the initial series of startup tests which included the critical experiment, zero and low-power tests, and pulse testing. The reactor design and results from the initial testing program are described in this paper. A portion of the startup experiments with SPR-III have been completed and this paper discusses the more important aspects of the initial testing program

  10. 78 FR 2678 - Proposed Collection; Comment Request (60-Day FRN): The National Cancer Institute (NCI...

    Science.gov (United States)

    2013-01-14

    ... Request (60-Day FRN): The National Cancer Institute (NCI) SmokefreeTXT (Text Message) Program Evaluation..., Behavioral Scientist/ Health Science Administrator, Division of Cancer Control and Population Sciences, 6130... text message smoking cessation intervention designed for young adult smokers ages 18-29. The Smokefree...

  11. Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

    In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

  12. Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

    International Nuclear Information System (INIS)

    Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

    2006-01-01

    Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

  13. Results of the initial test program for the Sandia Pulsed Reactor III (SPR III)

    International Nuclear Information System (INIS)

    Estes, B.F.; Reuscher, J.A.

    1976-08-01

    This document presents a detailed discussion of the reactor including the mechanical and nuclear design characteristics. Also presented are the complete results of the Initial Approach to Critical and the Zero-and-Low Power testing programs. Reactivity worth measurements are given for such parameters as control element integral worth, Safety Block integral worth, and various materials (polyethylene, copper, lead, etc) as a function of position relative to the core. Subcritical reactivity measurements made during the approach to critical generally proved to be in reasonably good agreement with design values due to the good source-fuel-detector geometry possible with a reactor of this type. Subsequent dynamic measurements for reactivity worths are shown to be in good agreement with calculated results

  14. Modernization of the facilities of the TRIGA Mark III reactor of ININ; Modernizacion de las instalaciones del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Mendez T, D.; Flores C, J., E-mail: dario.mendez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.

    2014-07-01

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

  16. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, H.; Emeterio H, M.; Canizal S, C. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, C.P. 11801 Mexico D.F. (Mexico)

    2000-07-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  17. Characterization of the TRIGA Mark III reactor for k0-neutron activation analysis

    International Nuclear Information System (INIS)

    Diaz R, O.; Herrera P, E.; Lopez R, M.C.

    1997-01-01

    The non-ideality of the epithermal neutron flux distribution in a a reactor site parameter (α), the thermal-to-epithermal neutron ratio (f), the irradiation channel neutron temperature (T n ) and the k 0 -factors for more than 20 isotopes were determined in the 3 typical irradiation positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using different experimental methods with conventional and non-conventional monitors. This characterization is used in the k 0 -method of NAA, recently introduced at the Institute. (author). 21 refs., 3 figs., 5 tabs

  18. Social value orientation modulates the FRN and P300 in the chicken game.

    Science.gov (United States)

    Wang, Yiwen; Kuhlman, D Michael; Roberts, Kathryn; Yuan, Bo; Zhang, Zhen; Zhang, Wei; Simons, Robert F

    2017-07-01

    Social dilemmas pervade daily life, business, and politics. The manners in which these dilemmas are resolved depend in part on the personal characteristics of those involved. One such characteristic is Social Value Orientation (SVO), a trait-like predisposition to maximize cooperative (Pro-Social) or non-cooperative (Pro-Self) outcomes in social relationships. The present study investigated the role of SVO in modulating neural responses to outcomes in a type of social dilemma known as the Chicken Game. The Chicken Game models real-world situations involving two parties independently making a decision between cooperation and aggression. The EEG of Pro-Socials and Pro-Selfs was recorded while playing Chicken with a computer Opponent. Two ERP components were extracted: Feedback-Related Negativity (FRN) and the P300. Despite no behavioral differences in decision (i.e., cooperation, aggression), FRN results indicate that Pro-Socials experienced unreciprocated cooperation as the least desired outcome. Further, P300 results show a main effect for the Opponent's choice, such that the Opponent's cooperation was more salient than aggression. Additionally, an interaction between the Participant's and Opponent's choice showed that the effect for the Opponent's choice only occurred when the Participant chose cooperation. None of the results for P300 were moderated by SVO. For both ERP components, Pro-Selfs showed no differential responding to Chicken outcomes. In addition, FRN magnitude on trial n predicted choice on trial n+1 for Pro-Socials, but not for Pro-Selfs. P300 magnitude on trial n showed no relationship to choice on trial n+1. Results indicate that individual differences in SVO modulate FRN responses to Chicken outcomes, and that these neural reactions may have utility in predicting subsequent behaviors. For P300, there is no evidence of SVO modulation. Our general pattern of FRN responsiveness in Pro-Socials, but not in Pro-Selfs, is related to similar findings in f

  19. The application and design of distributed control system in reactor shutdown system of Qinshan phase III

    International Nuclear Information System (INIS)

    Su Guoquan; Liu Wangtian; Yu Yijun; Xiong Weihua

    2006-03-01

    The design, commissioning and running of the reactor trip parameter monitoring system used in Qinshan Phase III are introduced. The applying technology of Distributed Control System realized trip parameter monitoring and realized the function of trip parameters quick data acquisitioning, transferring, saving, alarm, query. The applying of trip parameters monitoring system improved the abilities of plant status monitoring and event analyzing, and increased the security and economy of nuclear power plant. (authors)

  20. International Working Group on Past Reactors Thirteenth Annual Meeting. Summary Report. Part III

    International Nuclear Information System (INIS)

    1981-04-01

    The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

  1. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Method was developed for calculation of Xe 135 static effect and kinetic effects of Xe 135 and Sm 149 with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe 13 5 microscopic capture cross section dependent on temperature [sr

  2. Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Ramirez, R.M.; Arrendondo, R.R.

    1990-01-01

    The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235 U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

  3. Tritium system design for the mirror reactors FPD-I, FPD-II, and FPD-III

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

    The tritium system design for the Fusion Power Demonstration Reactor (FPD-I, II, and III) is described. The device operates at 25% availability. For FPD-II, an engineering mode using tritium neutral beams is part of the design

  4. Specialist committee's review reports for experimental fast reactor JOYO' MK-III performance tests

    International Nuclear Information System (INIS)

    Yamashita, Kiyonobu; Okubo, Toshiyuki; Kamide, Hideki

    2004-02-01

    Performance tests (startup-physics tests and power elevation tests) were planed for experimental fast reactor 'JOYO' MK-III where irradiation performances were upgraded by power increase from 100 to 140 MW. The reactor safety committee of O-arai Engineering Center has established a specialist committee for 'JOYO' MK-III Performance Tests at the first meeting of 2003 on 23th. April 2003, to accomplish the tests successfully. Subjects of the specialist committee were reviews of following items covering a wide range. 1) Contents of modification works. 2) Reflections of functional test results to the plant and facilities. 3) Reflections of safety rule modification to instruction and manual for operation. 4) Quality assurances and pre-calculation for performance test. 5) Inspection plan and its results. 6) Adequacy of performance test plan. 7) Confirmation of performance test results. Before test-starts, the specialist committee has confirmed by reviewing the items from 1) to 6) based on explanations and documents of the Division of Experimental Reactor, that the test plan and pre-inspections are adequate. After the tests, the specialist committee had confirmed by reviewing the item 7) in the same way, that the each test result satisfies the corresponding criterion. The specialist committee has concluded from these review's results before and after the tests that the 'JOYO' MK-III Performance Tests were carried out appropriately. Besides, the first criticality of the JOYO MK-III was achieved on 2nd. July 2003, and the continuous full power operation was carried on 20th. Nov. 2003. Finally, all performance tests were completed by the pass of the last governmental pre-serviced inspection (dose rate measurement during the shut down condition). (author)

  5. Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  6. Flow-induced vibration phenomenon in a Mark III TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C K; Whittemore, W L; Kim, B S; Lee, J B; Blevins, R D; Burton, T E [Korea Atomic Energy Research Institute, Seoul (Korea, Republic of); General Atomic Company, San Diego, CA (United States)

    1976-07-01

    The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

  7. Flow-induced vibration phenomenon in a Mark III TRIGA reactor

    International Nuclear Information System (INIS)

    Lee, C.K.; Whittemore, W.L.; Kim, B.S.; Lee, J.B.; Blevins, R.D.; Burton, T.E.

    1976-01-01

    The Mark III TRIGA reactor with hexagonal fuel spacing is capable of operating at 2.0 MW. The Mark III at San Diego operated without core cooling problems or vibration at power levels up to 2.0 MW. All Mark III reactors have operated trouble-free up to 1.0 MW. The Mark III TRIGA in Korea was installed in 1972 and operated many months without trouble at 2.0 MW. During this period core changes including addition of new fuel were made. Eighteen months after startup, a coolant flow-induced vibration was observed for the first time at a power of 1.5 MW. A lengthy series of tests showed that it was not possible to establish a core configuration that permitted vibration-free operation for power levels in the range 1.5 - 2.0 MW. Observations during the tests confirmed that standing waves in the reactor tank water coupled the source within the core to the shield structure and surrounding building. Analysis of the data indicates strongly that the source of the vibration is the creation and collapse of bubbles with the core acting as a resonator. A substantially increased flow of coolant through the upper grid plate is expected to eliminate the vibration phenomenon and permit trouble-free operation at power up to 2.0 MW. In an attempt to seek a remedy, both GAC and KAERI have independently developed designs for upper grid plates. KAERI has constructed and installed an interim version of the standard grid plate which was calculated to provide 25% more coolant flow and mounted high so as to provide less restriction to flow around the upper fittings of the fuel elements. A substantial reduction in vibration was observed. No vibration was observed at any power up to 2.0 MW with cooling water at or below 20 C. A slight vibration at 1.8 MW occurred for higher cooling temperatures. The GAC grid plate design provides not only for increasing the flow area but also for streamlining the flow surfaces on the grid plate and possibly also on the top fittings of the fuel elements. It is

  8. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mejia C, M. A.

    2016-01-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  9. Modernization of the facilities of the TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Mendez T, D.; Flores C, J.

    2016-09-01

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  10. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.

    2014-01-01

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

  11. Experimental measurement of the refrigerant temperature of the TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Gallardo S, L.F.; Alonso V, G.

    1991-08-01

    With the object of knowing the axial temperature profile of the refrigerant in the core of the TRIGA Mark III reactor of the ININ, the temperatures of this, at the enter, in the center and the exit of the core were measured, in the positions: west 2, north 2 and south 1. This was made by means of the thermo pars introduction mounted in aluminum guides, connected to a measurer of digital temperature, whose resolution is of ± 0.1 C. The measurements showed a bigger heating of the refrigerant in the superior half of the core, that which suggests that the axial profile of temperature of the reactor is not symmetrical with respect to the center or that those temperature measurements in the center are not correct. (Author)

  12. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  13. Safety in the ARIES-III D-3He tokamak reactor design

    International Nuclear Information System (INIS)

    Herring, J.S.; Dolan, T.J.

    1992-01-01

    This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

  14. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR; Diseno del mecanismo para la conduccion de las barras de control del reactor Triga Mark III del ININ.

    Energy Technology Data Exchange (ETDEWEB)

    Franco C, A

    1997-12-01

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author).

  15. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  16. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

    2009-10-01

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  17. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Garcia M, H.; Emeterio H, M.; Canizal S, C.

    1999-01-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  18. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  19. Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Larsen, E.M.

    1976-01-01

    Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed. (Auth.)

  20. Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

    1972-07-01

    This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

  1. Theoretical analysis of nuclear reactors (Phase III), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, III Deo, Zatrovanje reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Report on calculation of poisoning in experimental and power reactor includes four parts. Part one describes the influence of poisoning on the physical parameters of a reactor. part two includes transformation of differential equations for iodine and xenon. It was needed for easier solution of of differential equation using the analog computer. This calculation was done for RA reactor operating at 5 MW power. The RA reactor was used an example of calculation by the proposed method. Part four shows the application of the method for calculating the Calder Hall power reactor.

  2. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Martinez, H

    1999-07-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 {sup L}ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  3. Development and validation of a model TRIGA Mark III reactor with code MCNP5

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  4. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

    2013-10-01

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  5. Neutron detection of the Triga Mark III reactor, using nuclear track methodology

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

  6. Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

    International Nuclear Information System (INIS)

    Chen Yiping; Liu Songlin

    2010-01-01

    Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

  7. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

    2013-10-15

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  8. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  9. Summary report of the experimental fast reactor JOYO MK-III performance test

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  10. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

    2012-10-01

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

  11. Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

    1990-01-01

    It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

  12. Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

    1989-01-01

    The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

  13. Activation calculation of steel of the control rods of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A.

    2014-10-01

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  14. TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

    Science.gov (United States)

    Ferguson, Megan A; Hering, Janet G

    2006-07-01

    Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

  15. Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J L; Aguilar H, F; Rivero G, T; Sainz M, E [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

    2000-07-01

    In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

  16. 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).

  17. Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-02-01

    Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects

  18. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  19. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC; Coordinacion de los sistemas motrices de las barras de control del reactor Triga Mark III, mediante el uso de un PLC

    Energy Technology Data Exchange (ETDEWEB)

    Castro R, P. R.

    2016-07-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  20. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor; Desarrollo de la interface de usuario para la visualizacion de los sistemas auxiliares del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Merced D, J. E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  1. Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

    International Nuclear Information System (INIS)

    In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers

  2. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC

    International Nuclear Information System (INIS)

    Castro R, P. R.

    2016-01-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  3. Molten salt related extensions of the SIMMER-III code and its application for a burner reactor

    International Nuclear Information System (INIS)

    Wang Shisheng; Rineiski, Andrei; Maschek, Werner

    2006-01-01

    Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16-20 November 2003]. The molten salt fuel is a ternary NaF-LiF-BeF 2 system fuelled with ca. 1 mol% typical compositions of transuranium-trifluorides (PuF 3 , etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP' 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as

  4. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Merced D, J. E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  5. Estimation development cost, study case: Quality Management System Reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Antúnez Barbosa, Tereso Antonio; Valdovinos Rosas, Rosa María; Marcial Romero, José Raymundo; Ramos Corchado, Marco Antonio; Edgar Herrera Arriaga

    2016-01-01

    The process of estimating costs in software engineering is not a simple task, it must be addressed carefully to obtain an efficient strategy to solve problems associated with the effort, cost and time of activities that are performed in the development of an information system project. In this context the main goal for both developers and customers is the cost, since developers are worry about the effort pay-load and customers are worry about the product pay-load. However, in other fields the cost of goods depends on the activity or process that is performed, thereby deduce that the main cost of the final product of a development project software project is undoubtedly its size. In this paper a comparative study of common models for estimating costs are developed. These models are used today in order to create a structured analysis to provide the necessary information about cost, time and effort for making decisions in a software development project. Finally the models are applied to a case study, which is a system called Monitorizacion Automatica del Sistema de Gestion de Calidad del Reactor TRIGA Mark III. (author)

  6. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

    International Nuclear Information System (INIS)

    Garcia Martinez, H.

    1999-01-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  7. Behavior of exposed human lymphocytes to a neutron beam of the Reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Carbajal R, M. I.; Arceo M, C.; Aguilar H, F.; Guerrero C, C.

    2012-10-01

    The living beings are permanently exposed to radiations of natural origin: cosmic and geologic, as well as the artificial radiations that come from sources elaborated by the man. The artificial sources have an important use in the medical area. Particularly has been increased the neutrons use due to the effectiveness that they have to damage the cells with regard to other radiation types. The biological indicator of exposition to ionizing radiation more reliable is the chromosomal aberrations study, specifically the dicentrics in human lymphocytes. This test allows, establishing the exposition dose in function of the damage quantity. The dicentrics have a behavior in function of the dose. The calibration curve that describes this behavior is specific for each type of ionizing radiation. In the year 2006 beginning was given to the expositions of human lymphocytes to a neutron beam generated in the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico. Up to 2008 the response dose curve comprised an interval of exposition time of up to 30 minutes. Moreover, the interval between 10 an 20 minutes is included, since was observed that this last is indispensable for the adjustment waited in a lineal model. (Author)

  8. Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Herrera Peraza, E.

    1996-01-01

    The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

  9. Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2003-07-01

    The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

  10. Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

    International Nuclear Information System (INIS)

    Suzuki, Tohru; Tobita, Yoshiharu; Kondo, Satoru; Saito, Yasushi; Mishima, Kaichiro

    2003-01-01

    SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas-liquid flows, such as an air-water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking. It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka-Ishii's correlation has improved the accuracy of SIMMER-III for gas-liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed. Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios

  11. iB1350 no.1. A generation III.7 reactor after the Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    Sato, Takashi; Matsumoto, Keiji; Hosomi, Kenji; Kojima, Yoshihiro; Taguchi, Keisuke

    2017-01-01

    iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design, it has a double cylinder RCCV (Mark W containment) and in-depth hybrid safety systems (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, an innovative core catcher (iCC), a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. While the conventional PCCS can never cool the S/P, the iPCCS can automatically cool the S/P directly even in a DBA LOCA. That makes it possible for the iB1350 to optimize the active safety systems for a DBA. Sato came up with several optimized configurations of the IDHS that are expected to achieve further cost reduction and enhance its reliability resulting from passive feature of the iPCCS. The IC/iPCCS pool has enough capacity for 7 day grace period. The IC/iPCCS heat exchangers, the core and the spent fuel pool are

  12. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor; Elaboracion de la configuracion y programacion del sistema de interbloqueos del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Mejia C, M. A.

    2016-07-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  13. Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Carbajal R, M. I.

    2012-01-01

    Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + αD + βD. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y=αD. The dose-response relationship is established in blood samples from the same donor, the coefficient α of the dose-response is Y = (0.3692 ± 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

  14. ABWR1. A Generation III.7 reactor after the Fukushima Daiichi accident

    International Nuclear Information System (INIS)

    Sato, Takashi; Matsumoto, Keiji; Kurosaki, Toshikazu; Taguchi, Keisuke

    2015-01-01

    iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. The NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most

  15. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  16. RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1984-07-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

  17. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jorge.benitez@inin.gob.mx

    2009-10-15

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  18. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-05-15

    This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

  19. Verification of performance of the power percentage channel for the TRIGA Mark III reactor; Verificacion del funcionamiento del canal del porciento de potencia del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C

    1991-10-15

    It was found that the response that gives the power percent channel is correct, given the positive results of the independent tests that were carried out to the gamma ionization chamber and the electronics associated to this channel. Regarding the gamma chamber, it was verified that the appropriate operation voltage is 800 V, and that for operations in stationary state to 1 MW during 2 h, presented maximum variations of 3%. Also it was determined that the degradation percentage in the sensitivity to the gamma radiation is 10.24%, because this chamber has not been changed since the reactor enters in operation at November 8, 1968 by what will be considered to short term the substitution of the same one due to the burnt that it presents. In connection with the electronics of the channel, it was simulated the response of the chamber for intervals of 6 h and in the 4 analyzed cases the response of the channel was lineal. (Author)

  20. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2006-07-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  1. Characterization of the irradiation facilities SINCA and SIRCA of the TRIGA Mark III reactor using the code MCNPX

    International Nuclear Information System (INIS)

    Delfin L, A.; Garcia M, T.; Lucatero, M. A.; Cruz G, H. S.; Gonzalez, J. A.; Vargas E, S.

    2011-11-01

    The commitment of changing fuels of high enrichment for fuels of low enrichment in the TRIGA Mark III reactor of the Nuclear Center of Mexico generates the necessity to know the distribution of the spectrum of the neutrons flux in the irradiation facilities like they are: the Pneumatic System of Capsules Irradiation and the Rotational System of Capsules Irradiation. Is very important for the experiments design as well as for the reactor safety to know the profiles of the neutrons flux and the spectrum that these maintain with the mixed core with which operates, to effect of conserving the same characteristics when the reactor core will be operated with fuel of low enrichment totally. Also, knowing the profiles of the neutrons flux, the reactor operators can optimize the irradiation conditions of the processed samples and likewise the users can select the irradiation positions more adaptable to their necessities. This work present the characterization of the neutron flux in the irradiation facilities SINCA and SIFCA, calculated with the code MCNPX. (Author)

  2. Preliminary studies for the upgrading of the ININ TRIGA MARK III reactor

    International Nuclear Information System (INIS)

    Aguilar, J.V.

    1984-01-01

    An increase on the demand of services has led to the -consideration of reactor upgrading, for which a project has been proposed. This project includes the modification of the reactor core, using FLIP type (Erbium loaded, 70% U-235 enriched) fuel elements and the upgrading of the control - console which implies a careful analysis of the Safety -considerations involved in the system modification. In this paper, a brief description is given of: The modifications intended; Pre-operational tests for the system components to be modified; The safety report needed for the licensing of the upgraded reactor; The project management and evaluation

  3. Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Sook; Lee, Byung Doo

    2007-11-15

    Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report.

  4. Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

    International Nuclear Information System (INIS)

    Kim, Hyun Sook; Lee, Byung Doo

    2007-11-01

    Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report

  5. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    1991-05-01

    This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

  6. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.

    2014-07-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

  7. MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report

    International Nuclear Information System (INIS)

    Trosman, H.G.; Lanning, D.D.; Harling, O.K.

    1994-08-01

    The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers

  8. Operation and maintenance of the RA reactor in 1964, I-II, Part I

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1964-12-01

    During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core

  9. Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

    1995-03-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  10. Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment; Simulador de la cinetica puntual de un reactor nuclear TRIGA Mark III con control difuso de potencia en un ambiente visual

    Energy Technology Data Exchange (ETDEWEB)

    Perez M, C

    2004-07-01

    The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

  11. Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  12. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.

    2014-01-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -10 to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or research projects. (Author)

  13. A wide-range medition system for TRIGA Mark III Reactor

    International Nuclear Information System (INIS)

    Vazquez R, M.

    1995-01-01

    The number of particles emitted by a fission reaction is proportional to the number of fissions generated in the reactor nucleus, what in turn are proportional to the power level of such reactor; this indicates that it is possible could measure the reactor power if the amount of neutrons is measured, or the density of the neutron flux. The power measurement in the reactor is necessary in order to have a control of the same. Several procedures for power measurement exists, one of them is achieved through neutron flux density measurement, that take place in the chain reaction by means of the flow density measurement. The easiest way in order to achieve the neutron flux density measurement is carry out by means of the employment of neutron detectors. To the exit of these detectors an electric pulse taks place every time that a neutron interacts with the sensitive detector part. The work here presented, concrete to the construction of a system of measurement of the nuclear power reactor, that is based on the neutron flux applying some techniques of the neutron noise analysis. (Author)

  14. RB research nuclear reactor, Annual report for 1989, I - III; Istrazivacki nukleani reaktor RB (Izvestaj o radu u 1989. godini), I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Pesic, M; Hadimahmutovic, N; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1989-12-15

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989.

  15. Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment

    International Nuclear Information System (INIS)

    Perez M, C.

    2004-01-01

    The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

  16. Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor; Determinacion del perfil de flujos en funcion de la potencia en el dedal central del Reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia F, A.

    2010-07-01

    The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E{sub cd} approx 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

  17. Determination of the subcadmium flux in the Triga Mark III reactor of the Nuclear Center in Mexico

    International Nuclear Information System (INIS)

    Rodiguez V, F.J.

    1975-01-01

    The determination of the subcadmium flux profile (neutrons with energy between 0 and 0.4 eV approximatley) in the core of the reactor Triga Mark III in the Nuclear Center of Mexico was made. The technique used for that purpose consists in placing alternativelly indium sheets covered and uncovered with cadmium and separated by fragments of aluminium 2.54cm wide in an aluminium tube whose external diameter reaches 0.635cm. This tube is introduced in each of the 21 axial irradiation places of the sheets the activity of gamma rays of 1.293 MeV of In-116m produced during the irradiation, is measured using a monochannel analyzer. We obtain graphs of the specific counting ratio vs position for the sheets uncovered and covered with cadmium; using quadratic interpolation we compute in each position the missing specific measuring ratio. The difference between the specific counting ratios with our cadmium and with cadmium multiplied by a constant gives the subcadmium flux in the z point. For irradiation at 10 watts, 1 Kilowatt and 1 megawatt the results show profiles which in general terms are in accordance with those which were obtained theoretically for a cylyndrical reactor using group theory. The greatest flow given by the manufacturer of the reactor is in the values range obtained experimentally.For the calculations, the method uses a FORTRAN IV program so that the determinations can be made routinely. (author)

  18. Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor

    International Nuclear Information System (INIS)

    Garcia F, A.

    2010-01-01

    The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E cd ∼ 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

  19. RB Research nuclear reactor, Annual report for 1994, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1994. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, beograd (Yugoslavia); Kocic, A; Ilic, R; Dasic, N; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Serbia)

    1994-12-15

    Report on RB reactor operation during 1994 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization as well as operation of the VAX-8250 computer.

  20. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

    2006-07-01

    The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  1. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

    2007-07-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  2. Phosphated in aluminium 6061-T651 used in the pool of the TRIGA Mark III nuclear reactor; Fosfatado en aluminio 6061-T651 utilizado en la tina del reactor nuclear TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Espinosa L, J.; Pena B, A.; Perez F, C.; Sanchez C, M.; Vite T, M.; Vite T, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    We characterized a phosphated coating used a substrate in aluminium 6061-T651, which is used in the container of the TRIGA Mark III nuclear reactor. Characterization was made using MEB and X-ray diffraction techniques. Coating application has the function to prevent the corrosion. Coating was probed to test adhesion in accordance with the Standard ASTM D-4541, and the corrosion process using a salt spray (fog) camera, in accordance with the Standard ASTM B-117, so as we could phosphate the welding cord. These experiences were obtained using a Deep cell. Results obtained are going to phosphate 'in situ' using a mobile device which was patented for the National Institute of Nuclear Research (ININ) in the Mexican Institute of Intellectual Property (INPI). (Author)

  3. Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B.

    2003-01-01

    The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

  4. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.; Luis L, M. A. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Division de Ciencias Basicas, Av. San Pablo No. 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico); Raya A, R.; Cruz G, H. S., E-mail: roberto.raya@inin.gob.mx [ININ, Departamento del Reactor, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10{sup -10} to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of {sup 197}Au, {sup 58}Ni, {sup 115}In, {sup 24}Mg, {sup 27}Al, {sup 58}Fe, {sup 59}Co and {sup 63}Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  5. Inspection with non destructive assay techniques of the aluminium coating of the TRIGA Mark III reactor vat; Inspeccion con tecnicas de ensayos no destructivos del recubrimiento de aluminio de la tina del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Reyes A, A.I.; Gonzalez M, A.; Castaneda J, G.; Rivera M, H.; Sandoval G, I. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    In June 2000, the Reactor Department assigned to the Scientific Research Direction of the National Institute of Nuclear Research requested to the Non-destructive Assays Laboratory (LEND), assigned to the Materials Science Management, the inspection and measurement of thickness of the aluminium coating (liner) of the TRIGA Mark III reactor vat with non-destructive assay techniques, due to that the aluminium coating is exposed mainly to undergo slimming on its back side due to corrosion phenomena. Activity that was able to be carried out from april until august 2001. It is worth pointing out that this type of inspection with these techniques was realized by first time. The non-destructive assays (NDA) are techniques which use indirect physical methods for inspecting the sanitation of components in process or in service, for detect lack of continuity or defects which affect their quality or usefulness. The application of those do not alter the physical, chemical, mechanical or dimensional properties of the part subject of inspection. The results of the application of the ultrasound inspection techniques, industrial radiography and penetrating liquids are presented. (Author)

  6. Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada reaktora RA - I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects.

  7. Phosphated in aluminium 6061-T651 used in the pool of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Aguilar H, F.; Espinosa L, J.; Pena B, A.; Perez F, C.; Sanchez C, M.; Vite T, M.; Vite T, J.

    2001-01-01

    We characterized a phosphated coating used a substrate in aluminium 6061-T651, which is used in the container of the TRIGA Mark III nuclear reactor. Characterization was made using MEB and X-ray diffraction techniques. Coating application has the function to prevent the corrosion. Coating was probed to test adhesion in accordance with the Standard ASTM D-4541, and the corrosion process using a salt spray (fog) camera, in accordance with the Standard ASTM B-117, so as we could phosphate the welding cord. These experiences were obtained using a Deep cell. Results obtained are going to phosphate 'in situ' using a mobile device which was patented for the National Institute of Nuclear Research (ININ) in the Mexican Institute of Intellectual Property (INPI). (Author)

  8. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F.

    2007-01-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  9. Determination of the fluence profile in three dimension for the thermal column of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Urena N, F.; Delfin L, A.; Garcia M, T.

    2006-01-01

    In this work the results of the dosimetric properties of the lithium carbonate are presented (detecting), before the thermal neutrons. The process consists on irradiating samples of lithium carbonate in the installation of the thermal column of the TRIGA Mark III reactor, with a controlled period and with time intervals of 20 hours of irradiation. It is necessary to mention that the detectors were placed in different internal positions of the thermal column. With the purpose of being used these results for future studies, like it is the fluence profile in the thermal column. To use the BNCT technique (Boron Neutron Capture Therapy). Which is a binary technique that requires the simultaneous presence of a neutron flux with appropriate energy and a neutron captor (10B), those which interacting to attack to the tumor cells without producing significant damage to the tissues when both agents are separated. (Author)

  10. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Hernandez, L.V.; Hernandez, J.A.

    2006-01-01

    The quality has not only become one of the essential requirements of the product but rather at the present time it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  11. RA reactor safety analysis I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada Reaktora RA I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    The objective of the maximum credible accident analysis was to determine the integral radiation doses in the vicinity of the reactor and in the environment. In case of RA reactor the maximum credible accident, meaning release of the fission products, would be caused by fuel elements meltdown. This analysis includes the following calculation results: activity of the fission products, volatility of the fission products, concentration of radioactive materials in the air, analysis of the accident environmental effects.

  12. The ARIES-III D-3He tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1991-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs

  13. Development of source term evaluation method for Korean Next Generation Reactor(III)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geon Jae; Park, Jin Baek; Lee, Yeong Il; Song, Min Cheonl; Lee, Ho Jin [Korea Advanced Institue of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    This project had investigated irradiation characteristics of MOX fuel method to predict nuclide concentration at primary and secondary coolant using a core containing 100% of all MOX fuel and development of source term evaluation tool. In this study, several prediction methods of source term are evaluated. Detailed contents of this project are : an evaluation of model for nuclear concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant using purely MOX fuel, suggestion of source term prediction method of NPP with a core using MOX fuel.

  14. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.; Luis L, M. A.; Raya A, R.; Cruz G, H. S.

    2013-10-01

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10 -10 to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of 197 Au, 58 Ni, 115 In, 24 Mg, 27 Al, 58 Fe, 59 Co and 63 Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  15. Inspection with non destructive assay techniques of the aluminium coating of the TRIGA Mark III reactor vat

    International Nuclear Information System (INIS)

    Reyes A, A.I.; Gonzalez M, A.; Castaneda J, G.; Rivera M, H.; Sandoval G, I.

    2001-01-01

    In June 2000, the Reactor Department assigned to the Scientific Research Direction of the National Institute of Nuclear Research requested to the Non-destructive Assays Laboratory (LEND), assigned to the Materials Science Management, the inspection and measurement of thickness of the aluminium coating (liner) of the TRIGA Mark III reactor vat with non-destructive assay techniques, due to that the aluminium coating is exposed mainly to undergo slimming on its back side due to corrosion phenomena. Activity that was able to be carried out from april until august 2001. It is worth pointing out that this type of inspection with these techniques was realized by first time. The non-destructive assays (NDA) are techniques which use indirect physical methods for inspecting the sanitation of components in process or in service, for detect lack of continuity or defects which affect their quality or usefulness. The application of those do not alter the physical, chemical, mechanical or dimensional properties of the part subject of inspection. The results of the application of the ultrasound inspection techniques, industrial radiography and penetrating liquids are presented. (Author)

  16. III. Co-electrodeposition/removal of copper and nickel in a spouted electrochemical reactor.

    Science.gov (United States)

    Grimshaw, Pengpeng; Calo, Joseph M; Hradil, George

    2011-07-11

    Results are presented of an investigation of co-electrodeposition of copper and nickel from acidic solution mixtures in a cylindrical spouted electrochemical reactor. The effects of solution pH, temperature, and applied current on metal removal/recovery rate, current efficiency, and corrosion of the deposited metals from the cathodic particles were examined under galvanostatic operation. The quantitative and qualitative behavior of co-electrodeposition of the two metals from their mixtures differs significantly from that of the individual single metal solutions. This is primarily attributed to the metal displacement reaction between Ni(0) and Cu(2+). This reaction effectively reduces copper corrosion, and amplifies that for nickel (at least at high concentrations). It also amplifies the separation of the deposition regimes of the two metals in time, which indicates that the recovery of each metal as a relatively pure deposit from the mixture is possible. It was also shown that nitrogen sparging considerably increases the observed net electrodeposition rates for both metals - considerably more so than from solutions with just the single metals alone. A numerical model of co-electrodeposition, corrosion, metal displacement, and mass transfer in the cylindrical spouted electrochemical reactor is presented that describes the behavior of the experimental copper and nickel removal data quite well.

  17. Experimental fusion power reactor conceptual design study. Final report. Volume III

    International Nuclear Information System (INIS)

    Baker, C.C.

    1976-12-01

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

  18. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  19. Experimental fast reactor JOYO MK-III functional test. Primary auxiliary cooling system test

    International Nuclear Information System (INIS)

    Karube, Koji; Akagi, Shinji; Terano, Toshihiro; Onuki, Osamu; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro

    2004-03-01

    This paper describes the results of primary auxiliary cooling system, which were done as a part of JOYO MK-III function test. The aim of the tests was to confirm the operational performance of primary auxiliary EMP and the protection system including siphon breaker of primary auxiliary cooling system. The items of the tests were: (Test No.): (Test item). 1) SKS-117: EMP start up test. 2) SKS-118-1: EMP start up test when pony motor running. 3) SKS-121: Function test of siphon breaker. The results of the tests satisfied the required performance, and demonstrated successful operation of primary auxiliary cooling system. (author)

  20. Operation and maintenance of the RA reactor in 1964, I-II, Part II; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, II Deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    This volume of the report contains the following 15 Annexes: Improvement of the fuel cycle economy (record No. 37009803 in INIS DB); Analysis of neutron flux increase in horizontal experimental channels of the RA reactor record No. 37005698 in INIS DB); Application of the critical system for determining the thermal neutron flux in a research reactor with central horizontal reflector ( record No. 37055005 in INIS DB); Determining the capacity of the RA reactor heat exchanger dependent on the coolant water temperature and flow; Operation of the RA reactor in forced regime; Analysis of the CEN-132 heavy water pumps failures at the RA reactor from decontamination till present; Modifications in the vacuum loop of the distillation system; Report on decontamination of the evaporator and cleaning of the condenser of the distillation system; Operation of reactor at nominal power with reduced D{sub 2}O circulation; Cooling of the RA reactor with reduced flow rate in the heavy water loop; Measurement of the heavy water level in the fuel channels of the RA reactor; Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964; Conclusions of the experts group at the meeting held on November 23 1964; After heat and the cooling problem after RA reactor shut-down; Measurement of noise and vibrations on the Ra reactor heavy water system; Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor; Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor; Study of the modifications in the synchronous generators, heavy water pumps and condenser batteries of the RA reactor.

  1. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico

    International Nuclear Information System (INIS)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J.

    2001-01-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  2. Spectrographic determination of impurities in ceramic materials for nuclear fusion reactors III. Analysis of magnesium oxide

    International Nuclear Information System (INIS)

    Melon, A. M.; Roca, M.; Rucandio, M. I.

    1992-01-01

    The determination of minor and trace elements in the magnesium oxide, considered as possible ceramic material in thermonuclear fusion reactors, has been studied. The concentration ranges are 0.1 - 0.3% for Ca, Si and Y, and at the ppm level for Al, Co, Cr, Fe, Hf, K, Li, Mn, Na, Ni, Se, Ti, V and Zr. Atomic emission spectroscopy with direct current are excitation and photographic detection has been employed. In order to eliminate the effect due to the differences in density between standards and samples, which are a source of errors, a chemical treatment of both is carried out. Likewise, for attaining conditions more suitable for the volatilization of certain impurities, these are determined with the sample in fluoride form. (Author) 11 refs

  3. Spectrographic determination of impurities in ceramic materials for nuclear fusion reactors III. Analysis of magnesium oxide

    International Nuclear Information System (INIS)

    Melon, A.M.; Roca, M.; Rucandio, M.I.

    1992-01-01

    The determination of minor and trace elements in the magnesium oxide, considered as possible ceramic material in thermonuclear fusion reactors, has been studied. The concentration ranges are 0.1 - 0.3 % for Ca, Si and Y, and at the ppm level for Al, Co, Cr, Fe, Hf, K, Li, Mn, Na, Ni, Sc, Ti, V and Zr. Atomic emission spectroscopy with direct current arc excitation and photographic detection has been employed. In order to eliminated the effect due to the differences in density between standards and samples, which are a source of errors, a chemical treatment of both is carried out. Likewise, for attaining conditions more suitable for the volatilization of certain impurities, these are determined with the sample in fluoride form. (author)

  4. Verification of performance of the power percentage channel for the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    1991-10-01

    It was found that the response that gives the power percent channel is correct, given the positive results of the independent tests that were carried out to the gamma ionization chamber and the electronics associated to this channel. Regarding the gamma chamber, it was verified that the appropriate operation voltage is 800 V, and that for operations in stationary state to 1 MW during 2 h, presented maximum variations of 3%. Also it was determined that the degradation percentage in the sensitivity to the gamma radiation is 10.24%, because this chamber has not been changed since the reactor enters in operation at November 8, 1968 by what will be considered to short term the substitution of the same one due to the burnt that it presents. In connection with the electronics of the channel, it was simulated the response of the chamber for intervals of 6 h and in the 4 analyzed cases the response of the channel was lineal. (Author)

  5. Determination of mean molecular weights in organic reactor coolants. III. Differential cryoscopy with thermoelectric thermometer

    International Nuclear Information System (INIS)

    Becerro, E.; Carreira, M.

    1968-01-01

    The solubility problems raised by some components of the polymeric residue of irradiated polyphenolic coolants, which make it necessary to operate with very small samples, have been solved by means of a differential cryoscopic technique using a thermoelectric thermometer (thermal) as sensitive element. The method is based on the direct measurement of the difference between the freezing points of the investigated solution and of a reference solution whose concentration may be changed at will. The change of Δ V (mV) versus c(molal) is linear, the equivalent point being determined either analytically or graphically depending on the required accuracy. The method has been tested by measurements on pure polyphenyls, using diphenyl ether as solvent. It has been also applied to the main prospective coolants for the DON reactor. Working with 10 2 molal solutions the accuracy is better than ± 2 per cent. (Author) 2 refs

  6. FLICA III. A digital computer program for thermal-hydraulic analysis of reactors and experimental loops

    International Nuclear Information System (INIS)

    Plas, Roger.

    1975-05-01

    This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles [fr

  7. RB research nuclear reactor - Annual report for 1986, I - III; Istrazivacki nuklearni reaktor RB (Izvestaj o radu u 1986. godini), I-III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1987-07-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986. Izvestaj pokazuje podatke o radu reaktora RB u toku 1986. godine, stanje reaktorske opreme, podatke o angazovanom osoblju na reaktoru i datoteku sa podacima o eksperimentalnoj i drugoj opremi reaktora RB. Sastoji se od 3 dela: tehnicki opis, pogon i rad reaktora, oprema postrojenja i njeno odrzavanje, koriscenje reaktora u 1986. godini.

  8. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Barranco R, F.

    2015-01-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  9. The Phase I/II BNCT Trials at the Brookhaven medical research reactor: Critical considerations

    International Nuclear Information System (INIS)

    Diaz, A.Z.

    2001-01-01

    A phase I/II clinical trial of boronophenylalanine-fructose (BPA-F) mediated boron neutron capture therapy (BNCT) for Glioblastoma Multiforme (GBM) was initiated at Brookhaven National Laboratory (BNL) in 1994. Many critical issues were considered during the design of the first of many sequential dose escalation protocols. These critical issues included patient selection criteria, boron delivery agent, dose limits to the normal brain, dose escalation schemes for both neutron exposure and boron dose, and fractionation. As the clinical protocols progressed and evaluation of the tolerance of the central nervous system (CNS) to BPA-mediated BNCT at the BMRR continued new specifications were adopted. Clinical data reflecting the progression of the protocols will be presented to illustrate the steps taken and the reasons behind their adoption. (author)

  10. Burnable poison calculations for Mk.III gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Gubbins, M E

    1971-02-15

    A method of calculating the reactivity and burn-up hisotry of a Mk.III GCR system containing burnable poisons has been described. The method allows for poison-fuel interaction. Using the method it has been shown that burn-up of the poison under a constant incident flux can give errors of the order of 1-2 niles. A calculation using the method described will take about 50% longer than a straightforward fuel burn-up calculation in the same number of groups. The multi-cell approach has a potential for handling greater geometrical complexity. It is intended to compare the method against experiment as soon as suitable experimental results become available.

  11. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    International Nuclear Information System (INIS)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

    2015-09-01

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  12. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  13. The Fort St. Vrain high temperature gas-cooled reactor. III

    International Nuclear Information System (INIS)

    Olson, H.G.; Brey, H.L.

    1979-01-01

    The helium circulator auxiliary system provides buffer helium and bearing water for the reactor's four circulators with two nearly identical auxiliary loops serving the two circulators of a primary coolant loop. A series of drains removes the water and helium for separation and recycle. Loss of buffer helium's function as a dynamic seal has resulted in inleakage of bearing water into the primary coolant and outleakage of primary coolant into the auxiliary system. Inleakage of water also has occurred due to inadvertent pressurization of the bearing cavity with the static shutdown seal set. Satisfactory performance of the normal, backup and emergency bearing water systems has been accomplished after numerous component additions and modifications. Frequent circulator trips have occurred. Most of these have involved the delicate sensors that measure buffer helium differential pressure. Transients in one loop have communicated to the other loop through common components. Total separation of the auxiliary loops will occur after the planned installation of those components that currently service both loops. (Auth.)

  14. Behavior of exposed human lymphocytes to a neutron beam of the Reactor TRIGA Mark III; Comportamiento de linfocitos humanos expuestos a un haz de neutrones del Reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Carbajal R, M. I.; Arceo M, C.; Aguilar H, F.; Guerrero C, C., E-mail: citlali.guerrero@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The living beings are permanently exposed to radiations of natural origin: cosmic and geologic, as well as the artificial radiations that come from sources elaborated by the man. The artificial sources have an important use in the medical area. Particularly has been increased the neutrons use due to the effectiveness that they have to damage the cells with regard to other radiation types. The biological indicator of exposition to ionizing radiation more reliable is the chromosomal aberrations study, specifically the dicentrics in human lymphocytes. This test allows, establishing the exposition dose in function of the damage quantity. The dicentrics have a behavior in function of the dose. The calibration curve that describes this behavior is specific for each type of ionizing radiation. In the year 2006 beginning was given to the expositions of human lymphocytes to a neutron beam generated in the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico. Up to 2008 the response dose curve comprised an interval of exposition time of up to 30 minutes. Moreover, the interval between 10 an 20 minutes is included, since was observed that this last is indispensable for the adjustment waited in a lineal model. (Author)

  15. Corrosion in the aluminum containment tank at the Nuclear Center of Mexico TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mota, Juan Ramon

    1986-01-01

    The reactor developed a leak inside the exposure room discovered when it was opened for a routine inspection. This leak started to diminish immediately after it was found and disappeared completely in 2.5 months. The hydrostatic tests of the exposure room cooling water pipes and of the primary cooling system suction pipe proved that piping do not have leaks. A portion of the total volume of water was drained from the pool to conduct an inspection on the aluminum liner. Penetrant dye tests were initiated over welded Joints and walls. Welded Joints were all found to be in good condition but a total of 35 indications were reported on walls and concentrated on two main areas. A vacuum system was used to test for leakage. Seven indications were found to be perforations that crossed through the wall, fifteen indications did not cross through the wall but required repair and the rest were superficial irregularities. For the inspection of surfaces that remained covered by water, two methods were used. One was a television camera that was adapted to be used under water and hooked to a monitor and a videorecorder for close up inspection of the walls. The other consisted of submarine still color photography performed by divers. The evaluation of these inspections concluded that out of the 10 areas previously identified, only one presented the kind of problem that required repair. The last inspection performed was that using ultrasound techniques. Irregularities found did not require complete replacement of the aluminum liner. The repair procedures included the welding of aluminum plates over damaged areas and the injection of an effective insulating material (resin) to stop the corrosion mechanism

  16. Visual inspection system and sipping design for spent fuel at TRIGA MARK III reactor of Mexico

    International Nuclear Information System (INIS)

    Delfin, A.; Mazon, R.

    2002-01-01

    In the framework of the Technical Cooperation Regional Project for Latin America RLA/4/018 for the biennium 2001-2002, one of the activities identified is the characterization of spent fuel. Of these activities an important one is not doubt the physical condition of spent fuel because an appropriate identification of the fuel status will prevent problems of fuel leaks, corrosion problems etc. As part of the activities of the project was decided that countries no having visual inspection and sipping systems should be very desirable to have them as a result of this project. The Triga reactor of Mexico does not have both of them, therefore, it was decided the need of having both system. The paper describe first the way we designed and constructed a remote Visual Inspection System and example of how is operated. Along the experience and problems we have had with the system. Also we will present the design of the Sipping system were two option were considered. First to take a sample of water after a convenient period of time passing through a circuit to a multichannel analyzer and to identify leakage by way of measuring Caesium-137. Second, exists the possibility that the Stainless Steel sleeve of the fuel has only very small failures, so it is going to be very difficult to have leakages unless the fuel is hot. Therefore we are evaluating the possibility of using heaters to increase the temperature of the fuel and succeed on detecting leakages. The results - we hope - will be ready to be presented at the meeting. (author)

  17. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  18. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  19. Operation and maintenance of the RA reactor in 1964, I-II, Part I; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, I Deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core.

  20. CCF analysis of BWR reactor shutdown systems based on the operating experience at the TVO I/II in 1981-1993

    International Nuclear Information System (INIS)

    Mankamo, T.

    1996-04-01

    The work constitutes a part of the project conducted within the research program of the Swedish Nuclear Power Inspectorate SKI, aimed to develop the methods and data base for the Common Cause Failure (CCF) analysis of highly redundant reactor scram systems. The data analysis for the TVO I/II plant is focused on the hydraulic scram system, and control rods and drives. It covers operating experiences from 1981 through 1993. (9 refs., 9 figs., 7 tabs.)

  1. RA reactor safety analysis, Part II - Accident analysis; Analiza sigurnosti rada Reaktora RA I-III, Deo II - Analiza akcidenta

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Radanovic, Lj; Milovanovic, M; Afgan, N; Kulundzic, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    This part of the RA reactor safety analysis includes analysis of possible accidents caused by failures of the reactor devices and errors during reactor operation. Two types of accidents are analyzed: accidents resulting from uncontrolled reactivity increase, and accidents caused by interruption of cooling.

  2. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  3. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Delfin L, A.; Palacios, J.C.; Alonso, G.

    2006-01-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  4. Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

  5. Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) to more..

    Science.gov (United States)

    Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) more...

  6. Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

    Science.gov (United States)

    Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

  7. Safety analysis of RA Reactor operation I-III; Analiza sigurnosti rada Reaktora RA I - III, IZ-213-0322-1963

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    This safety analysis report covers the following three parts: Technical and operational characteristics of the RA reactor; Accidents analysis; and Environmental effects of the maximum possible accident. [Serbo-Croat] Ovaj izvestaj o analizi sigurnosti rada reaktora RA sastoji se od tri dela: Tehnicke i pogonske karakteristike reaktora RA; Analiza akcidenta; i Posledice maksimalno moguceg akcidenta na okolinu reaktora.

  8. Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III; Comportamiento de linfocitos humanos expuestos a un haz de neutrones del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Carbajal R, M. I.

    2012-07-01

    Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + {alpha}D + {beta}D. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y={alpha}D. The dose-response relationship is established in blood samples from the same donor, the coefficient {alpha} of the dose-response is Y = (0.3692 {+-} 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

  9. Theoretical analysis of nuclear reactors (Phase II), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, III Deo, Zatrovanje reaktora, II faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-10-15

    This phase is dealing with influence of all the fission products except Xe{sup 135} on the reactivity of a reactor, usually named as reactor poisoning. The first part of the report is a review of methods for calculation of reactor poisoning. The second part shows the most frequently used method for calculation of cross sections and yields of pseudo products (for thermal neutrons). The system of equations was adopted dependent on the conditions of the available computer system. It is described in part three. Detailed method for their application is described in part four and results obtained are presented in part five.

  10. Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Diaz G, C. A.

    2016-01-01

    Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm"2/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

  11. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  12. Pre-Preliminary results from the phase III of the IAEA CRP: optimizing of reactor pressure vessel surveillance programmes and their analysis

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M; Gillemot, F; Kryukov, A; Levit, V

    1994-12-31

    This paper gives preliminary results and some conclusions from Phase III of the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses`` carried out during the last seven years in 15 member states. First analysis concerned: comparison of results from initial, un-irradiated materials condition, comparison of transition temperature shifts (from notch toughness testing) with respect to content of residual (P, Cu) and alloying (Ni) elements, type of material (base and weld metal), irradiation temperature (288 and 265 C), and type of fluence dependence. Special effort has been taken to the analysis of the behaviour of a chosen reference steel. (JRQ). 6 figs., 4 tabs.

  13. The development of the nuclear physics in Latvia III. The research nuclear reactor IRT begins to work in Latvia

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2005-01-01

    This article is associated with the study of reactors technical parameters with specific interest on the effect the distribution of neutron and gamma radiation through the reactor's cooling systems has on the environment. Scientist began by implementing monitoring system to assist in the research of nuclear spectroscopy, neutron activation analysis, neutron diffraction, solid-state radiation physics, chemistry and radiobiology. The first sets of results are summarized with in the article. (author)

  14. Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor; Desarrollo de un simulador para diseno y prueba de controladores de potencia en un reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Perez M, C.; Benitez R, J.S.; Lopez C, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

  15. Radiation protection at the RA reactor in 1984, Part III Removal of the liquid radioactive effluents for the needs of the RA reactor

    International Nuclear Information System (INIS)

    Mandic, M.; Plecas, I.; Vukovic, Z.; Knezevic, Lj.; Jankovic, O.; Kostadinovic, A.; Mihailovic, B.

    1984-01-01

    Contaminated water originates from: hot cells, heavy water distillation device, storage pools for cooling and cutting of fuel elements, water biological shield of the reactor. During 1984, 400 liters of water contaminated by 60 Co was treated. Most recent measurements showed that the VR-1 pool contains 280 m 3 of effluents having specific activity of 3.3 10 4 Bq/ml, and VR-2 contains 30 m 3 with specific activity of 4 10 3 Bq/ml

  16. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  17. Subcadmic and epicadmic flow in the dry tube of the TRIGA Mark III reactor of the Nuclear Center of Mexico

    International Nuclear Information System (INIS)

    Delfin L, A.; Mazon R, R.; Nava R, B.

    1991-04-01

    The mensuration of the thermal and fast flows of the irradiation facilities of the core of the reactor is important, since allow us to determine the optimum time of irradiation of the samples in the reactor. The Dry Tube especially, is an irradiation installation that it was designed in the I.N.I.N. to supply the pneumatic irradiation system of capsules with durations bigger than 15 minutes and it can be used for exposures until a maximum of three hours. The main users are the Nuclear Chemistry Department and the Neutron activation analysis. In this report the neutron flux sub cadmic and epi cadmic obtained in an experimental way in the Dry Tube for the reactor operating in stationary state to powers of 100 Kw, 300 Kw and 1000 Kw are reported and with these values it is interpolated for other powers. (Author)

  18. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Celestino M, E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  19. Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Perez M, C.; Benitez R, J.S.; Lopez C, R.

    2003-01-01

    The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

  20. Reactor building pressure proof test (PPT) and leak rate test (LRT) of Qinshan phase III (CANDU) project

    International Nuclear Information System (INIS)

    Gu Jun; Shi Jinqi; Fan Fuping

    2004-12-01

    As the first reactor building (R/B) without stainless steel liner in china, TQNPC studied the containment characteristics, such as strong concrete absorb/release air effect, poor containment penetration. etc. And carefully prepared test scheme and emergency response, creatively introduced the instrument air self-supply system in reactor building, developed the special measurement and analysis system for PPT and LRT, organized work under high-pressure on large-scale in the test. Finally got the containment leak rate result and the test-cost-time value is the best in all same type tests. (authors)

  1. Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

    1992-01-01

    This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

  2. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

    International Nuclear Information System (INIS)

    Hernandez G, J.

    2012-10-01

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  3. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  4. An assessment of BWR [boiling water reactor] Mark III containment challenges, failure modes, and potential improvements in performance

    International Nuclear Information System (INIS)

    Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J.

    1991-01-01

    This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs

  5. Theoretical evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    1991-11-01

    It was theoretically determined the accumulation of the Xe 135 and Sm 149 in function, of the time during a stationary state of 72 h. continuous for the reactor TRIGA Mark III to 1 MW of thermal power with mixed core. The values of negative reactivity due to these isotopes are of 2.04 dollars and 0.694 dollars to the 72 h, quantities that will have to be compensated if wants that the reactor continues working to this power. Under the same conditions but considering a core with standard fuel, it was found a value of ρ = 1.70 dollars, resulting a difference of 0.30 dollars of negative reactivity in function of the type of analyzed core. This difference is important for the calculations of fuel management of a reactor. The concentration in balance of the xenon was reaches after an operation to constant power of 1 MW by 50 h, contrary to the samarium that reaches it balance after 3 weeks of operation starting from the initial start up and it stays constant along the useful life of the reactor while a change of fuel doesn't exist. It was obtained that for operation times greater to 60 h. at 1 MW, a peak of negative reactivity of the Xe 135 is generated between the 7 and 11 h after the instantaneous shut down, with a value of 2.43 dollars, that is to say 0.39 additional dollars to those taken place during the continuous irradiation. (Author)

  6. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi; Ishiguro, Okikazu; Kuroki, Syuzi

    1977-10-01

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  7. The near-threshold high R-ratio fatigue crack growth characteristics of SA508 cl III reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Achilles, R.D.; Bulloch, J.H.

    1989-01-01

    This paper describes the effect of frequency and environment on the near-threshold fatigue crack growth behaviour of SA508 cl III reactor pressure vessel (RPV) steel. The study has shown that in the near-threshold regime microstructure and environment markedly affect fatigue crack growth behaviour. In an aqueous environment, fatigue crack growth behaviour became even more sensitive to microstructure, and the fatigue crack growth rate increased by a factor of four in the case of the 3Hz test, while that for the 0.3Hz test was increased by a factor of approximately sixteen. This environmental enhancement manifested itself in the form of intergranular failure. For the 0.3Hz test the percentage intergranular failure decreased from 18% to <1% with an increase in ΔK level. The transition from microstructure-sensitive to microstructure-insensitive occurs when the cyclic plastic zone size is of the order of the prior austenite grain size. (author)

  8. Support to the elaboration of the engineering of detail, configuration and programming of the control system of heat removal of the TRIGA Mark III reactor; Apoyo a la elaboracion de la ingenieria de detalle, configuracion y programacion del sistema de control de remocion de calor del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Diaz G, C. A.

    2016-07-01

    Nowadays, the peaceful and responsible use of nuclear energy in Mexico is of great importance and contributes to economic, social, scientist and technologic development in the country, highlighting the Instituto Nacional de Investigaciones Nucleares (ININ) and the Nuclear Power Plant of Laguna Verde as one of the most important dependences. Among the main facilities and laboratories of ININ is the Nuclear Research Reactor TRIGA Mark III, this is a pool type reactor with mobile core, cooled and moderated by light water and a flow of 1013 n/cm{sup 2}/sec. Due to the technological obsolescence is a growing problem that threatens the information, operation and/or efficacy of elements of control and safety systems of the reactor, these must be changed each time more frequently. In the modernization of reactor was used a Modicon M340 programmable logic control (PLC) and a Twido PLC for the control of heat removal system (Primary Cooling System (PCS) and Secondary Cooling System (SCS) respectively), this because the PLC has proven to be safe and effective devices, addition to reduce the wiring elements and increase the possibilities of performance and design of the digital control console. This document shows and describes the elements of heat removal system (PCS and SCS), and the signals and signal types that such items send or received by the PLC, likewise, is indicated the methodology used to develop the applications for the control of the Primary Cooling System and Secondary Cooling System, beginning with the PLC design, the development of PLC plans and the control logic, and finally, the simulation and debugging of applications on Unity Pro and Twido Suite. All this in compliance with the safety standards to nuclear research reactors (NS-R-4), the rules of industrial programming (IEC 61131-3), and the reactor operating limits postulated in the safety report and the software assurance system used in the ININ. (Author)

  9. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, III Deo, Zatrovanje reaktora, I faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Method was developed for calculation of Xe{sup 135} static effect and kinetic effects of Xe{sup 135} and Sm{sup 149} with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe{sup 135} microscopic capture cross section dependent on temperature. Razradjen je metod proracuna statickog efekta Xe{sup 135} zatim kinetickog efekta Xe{sup 135} i Sm{sup 149} sa posebnim tretiranjem jodne jame i promene zatrovanja pri prelazu sa jedne snage na drugu. Navedeni efekti su tretirani prvo za uransko gorivo, a zatim su glavni obrasci uopsteni za smesu fisibilnih materijala. U prilogu su dati u vidu tabele, podaci potrebni za proracun i grafik zavisnosti mikroskopskog preseka zahvata Xe{sup 135} od temperature.

  10. Assessment of very high-temperature reactors in process applications. Appendix III. Engineering evaluation of process heat applications for very-high temperature reactors

    International Nuclear Information System (INIS)

    Wiggins, D.S.; Williams, J.J.

    1977-04-01

    An engineering and economic evaluation is made of coal conversion processes that can be coupled to a very high-temperature nuclear reactor heat source. The basic system developed by General Atomic/Stone and Webster (GA/S and W) is similar to the H-coal process developed by Hydrocarbon Research, Inc., but is modified to accommodate a nuclear heat source and to produce synthetic natural gas (SNG), synthesis gas, and hydrogen in addition to synthetic crude liquids. The synthetic crude liquid production is analyzed by using the GA/S and W process coupled to either a nuclear- or fossil-heat source. Four other processes are included for comparison: (1) the Lurgi process for production of SNG, (2) the Koppers-Totzek process for production of either hydrogen or synthesis gas, (3) the Hygas process for production of SNG, and (4) the Westinghouse thermal-chemical water splitting process for production of hydrogen. The production of methanol and iron ore reduction are evaluated as two potential applications of synthesis gas from either the GA/S and W or Koppers-Totzek processes. The results indicate that the product costs for each of the gasification and liquefaction processes did not differ significantly, with the exception that the unproven Hygas process was cheaper and the Westinghouse process considerably more expensive than the others

  11. Influence of the control bars pattern on the response of the operation channels of the TRIGA Mark III reactor; Influencia del patron de barras de control sobre la respuesta de los canales de operacion del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-07-15

    The local flow perturbations not generated by movements of bars not planned adequately to operate the reactor to 1 MW of thermal power, are reflected in the independent responses of the operation channels of the same one, find variations average from 17% to 30% for the channel of the power percent and of until 10% for the logarithmic channel. For the case of the lineal and percent power channels, these are between 14% and 46% as maximum when moving some of the bars. These variations can diminish until 5% in the channel of the power percent and until 3% on the average for the logarithmic one, all times when the calculated bars pattern for that irradiation considers that all the bars operate inside the lineal region of its calibration curve with approximately the same reactivity value each one and that during the operation the required reactivity compensations are carried out with the diametrically opposed bar to the irradiation installation used in that experiment. (Author)

  12. Tasks related to increase of RA reactor exploitation and experimental potential, 04. Device for transport of radioactive reactor channels and semi channels of the RA reactor, design project (I-III) Part II, Vol. II; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 04. Uredjaj za transport aktivnih tehnoloskih kanala I semikanala reaktora RA - izrada projekta (I-III), II Deo, Album II

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    This second volume includes calculations of the main components of the transporter, description of the mechanical part of the transporter and the engineering drawing of the device for transport of radioactive reactor channels and semi channels of the RA reactor.

  13. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ; Diseno, construccion e implementacion de dos circuitos redundantes de la logica de actuacion del sistema de proteccion de la nueva consola de control del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Celestino M, E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  14. Experimental evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core, configuration No. 16 (Final report of the project); Evaluacion experimental de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto, config. No. 16 (Informe final del proyecto)

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C

    1991-11-15

    It was generated the concentration curve of the Xe{sup 135} (t) during the TRIGA Mark III reactor operation cycle, for a continuous irradiation of 72 h to 1 MW of thermal power, as well as the accumulation curve of the isotope after the shutdown, for the fuel configuration No. 16 in the thermal column. The maximum negative reactivities generated by the Xe{sup 135} for operation times greater than 60 h to 1 MW and after the reactor shutdown its were of 1.968 {+-} 0.15 dollars and 2.30 {+-} 0.15 dollars respectively. When comparing these results with those theoretically calculated we find differences of the order of 3.6% and 5.34% which are understood inside the experimental error that on the average was of 7.6%. The results before mentioned have an important application during the start up process of the Reactor, when analyzing the value of the weekly reactivity excess of the core and when is choice the pattern of bars to use for experiments of but of 2 h, where is required to minimize the temporary and space interferences of the neutron flux. (Author)

  15. Influence of the control bars pattern on the response of the operation channels of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    1991-07-01

    The local flow perturbations not generated by movements of bars not planned adequately to operate the reactor to 1 MW of thermal power, are reflected in the independent responses of the operation channels of the same one, find variations average from 17% to 30% for the channel of the power percent and of until 10% for the logarithmic channel. For the case of the lineal and percent power channels, these are between 14% and 46% as maximum when moving some of the bars. These variations can diminish until 5% in the channel of the power percent and until 3% on the average for the logarithmic one, all times when the calculated bars pattern for that irradiation considers that all the bars operate inside the lineal region of its calibration curve with approximately the same reactivity value each one and that during the operation the required reactivity compensations are carried out with the diametrically opposed bar to the irradiation installation used in that experiment. (Author)

  16. Theoretical evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core; Evaluacion teorica de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-11-15

    It was theoretically determined the accumulation of the Xe{sup 135} and Sm {sup 149} in function, of the time during a stationary state of 72 h. continuous for the reactor TRIGA Mark III to 1 MW of thermal power with mixed core. The values of negative reactivity due to these isotopes are of 2.04 dollars and 0.694 dollars to the 72 h, quantities that will have to be compensated if wants that the reactor continues working to this power. Under the same conditions but considering a core with standard fuel, it was found a value of {rho} = 1.70 dollars, resulting a difference of 0.30 dollars of negative reactivity in function of the type of analyzed core. This difference is important for the calculations of fuel management of a reactor. The concentration in balance of the xenon was reaches after an operation to constant power of 1 MW by 50 h, contrary to the samarium that reaches it balance after 3 weeks of operation starting from the initial start up and it stays constant along the useful life of the reactor while a change of fuel doesn't exist. It was obtained that for operation times greater to 60 h. at 1 MW, a peak of negative reactivity of the Xe{sup 135} is generated between the 7 and 11 h after the instantaneous shut down, with a value of 2.43 dollars, that is to say 0.39 additional dollars to those taken place during the continuous irradiation. (Author)

  17. Present status of reactor physics in the United States and Japan-III. 2. Nuclear Fuel Management Optimization Capabilities

    International Nuclear Information System (INIS)

    Karve, Atul A.; Keller, Paul M.; Turinsky, Paul J.; Maldonado, G. Ivan

    2001-01-01

    approach is that many histories, i.e., combinations of decision variables, must be evaluated, which implies that many core physics calculations are required to determine the family of near-optimum decisions. To reduce computer execution time, highly efficient, core physics models with only the fidelity required for the assigned task are utilized. Ideally, one would like to utilize the same core physics models for all nuclear problems for consistency and ease of usage, which may someday occur with increases in computational power and advances in computational reactor physics. In considering core physics models, a unique aspect for nuclear fuel management optimization applications is that many repetitive calculations need to be completed during the optimization search. This implies that considerable overhead can be tolerated to reduce the computational time per history since the overhead will be amortized over many histories. This feature can imply the employment of different solution approaches than normally utilized. How various suboptimum problems integrate in an attempt to address the global optimization problem is now explained. The out-of-core optimization OCEON-P code has a number of decision variables, but the only decision that carries-forward in the reload design process is the cycling scheme, i.e., batch sizes in each cycle of the planning horizon. Note that OCEON-P is the only optimization code within the suite that truly does multicycle optimization and so can meaningfully evaluate and minimize levelized fuel cycle cost. The FORMOSA-L code optimizes the lattice, normally constrained to follow a specified reactivity versus burnup. This constraint provides the linkage to the core-wide analysis but is problematic to obtain. There currently does not exist within the suite of codes one that addresses the suboptimum problem of bundle design, which other researchers have addressed to a limited extent. With our current capabilities, a number of different bundle designs

  18. RA Research reactor, Part I: Technical and operational properties of the RA reactor; Analiza sigurnosti rada Reaktora RA I-III, Deo I: Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Zecevic, V; Nikolic, M; Popovic, B; Milosevic, M; Milic, M; Strugar, P; Pesic, M; Nikolic, V; Rajic, M; Radivojevic, J; Jankovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    RA reactor is a research reactor with rather high power density. Apart from research it is used for isotope production and industrial applications due to high reactivity excess (about 11%). It is a thermal reactor, heavy water moderated, cooled by D{sub 2}O, and H{sub 2}O, with a graphite reflector. Nominal power is 6.5 MW. Fuel is 2% enriched metal uranium, reactor core height is 1220 mm, and diameter is 1405 mm. Reactor lattice is square with lattice pitch 130 mm. There is 6 horizontal experimental channels and a graphite column. There is a total of 84 fuel channels and 45 experimental channels in the core. Maximum thermal neutron flux is 5.5 10{sup 13} n/cm{sup 2} s at nominal power level.

  19. Activities needed for exploitation of the RA reactor I-IV, Part III, IZ-093-0103-1961; Radovi za potrebe eksploatacije raktora RA - I-IV - III deo, IZ-093-0103-1961

    Energy Technology Data Exchange (ETDEWEB)

    Jurida, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    For the purpose of designing the low-temperature loop at the RA reactor, a list of materials that would be irradiated in the reactor was made. This document includes a review of the radiochemistry data for the listed materials and data about thermal stability of the mentioned materials. Calculation of heat generated in listed samples was based on the theoretical and experimental results of radiation characteristics of the RA reactor. Maximum temperature values in the samples without forced cooling are calculated. [Serbo-Croat] Za potrebe projektovanja niskotemperaturne petlje na reaktoru RA nacinjen je pregled materijala za ozracivanje u reaktoru. Ovaj dokument sadrzi pregled radiajaciono hemijskih podataka za navedene materijale i podatke o termickoj stabilnosti pomenutih materijala. Toplota generisana u navedenim uzorcima racunata je uzimajuci u obzir osnovne podatke teorijske i eksperimentalne rezultate odredjivanja karakteristika zracenja u reaktoru. Izracunate su maksimalne temperature u uzorcima bez prinudnog hladjenja.

  20. RA Reactor operation and maintenance (I-IX), Part III, Task 3.08/04-02 Refurbishment of the electrical equipment; Pogon i odrzavanje reaktora RA (I-IX), III Deo, Zadatak 3.08/04-02 Remont elektro opreme

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M; Poznanovic, B; Rajic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This volume contains detailed action plan for repair of electrical equipment of the RA reactor, the list of electrical equipment parts which were either repaired or exchanged for improvement of their performance. Detailed work describing the repair and maintenance work done of the listed equipment is part of this report. Equipment related to dosimetry and control systems are included as well.

  1. Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings

  2. Determination of mean molecular weights in organic reactor coolants. III. Differential cryoscopy with thermoelectric thermometer; Determinacion de masas moleculares medias en refrigerantes nucleares organicos. III-Crioscopia diferencial con termometro termoelectrico

    Energy Technology Data Exchange (ETDEWEB)

    Becerro, E; Carreira, M

    1968-07-01

    The solubility problems raised by some components of the polymeric residue of irradiated polyphenolic coolants, which make it necessary to operate with very small samples, have been solved by means of a differential cryoscopic technique using a thermoelectric thermometer (thermal) as sensitive element. The method is based on the direct measurement of the difference between the freezing points of the investigated solution and of a reference solution whose concentration may be changed at will. The change of {delta} V (mV) versus c(molal) is linear, the equivalent point being determined either analytically or graphically depending on the required accuracy. The method has been tested by measurements on pure polyphenyls, using diphenyl ether as solvent. It has been also applied to the main prospective coolants for the DON reactor. Working with 10{sup 2} molal solutions the accuracy is better than {+-} 2 per cent. (Author) 2 refs.

  3. Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

    International Nuclear Information System (INIS)

    Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

    2010-01-01

    The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison with some

  4. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  5. Measurement of europium (III)/europium (II) couple in fluoride molten salt for redox control in a molten salt reactor concept

    Science.gov (United States)

    Guo, Shaoqiang; Shay, Nikolas; Wang, Yafei; Zhou, Wentao; Zhang, Jinsuo

    2017-12-01

    The fluoride molten salt such as FLiNaK and FLiBe is one of the coolant candidates for the next generation nuclear reactor concepts, for example, the fluoride salt cooled high temperature reactor (FHR). For mitigating corrosion of structural materials in molten fluoride salt, the redox condition of the salts needs to be monitored and controlled. This study investigates the feasibility of applying the Eu3+/Eu2+ couple for redox control. Cyclic voltammetry measurements of the Eu3+/Eu2+ couple were able to obtain the concentrations ratio of Eu3+/Eu2+ in the melt. Additionally, the formal standard potential of Eu3+/Eu2+ was characterized over the FHR's operating temperatures allowing for the application of the Nernst equation to establish a Eu3+/Eu2+ concentration ratio below 0.05 to prevent corrosion of candidate structural materials. A platinum quasi-reference electrode with potential calibrated by potassium reduction potential is shown as reliable for the redox potential measurement. These results show that the Eu3+/Eu2+ couple is a feasible redox buffering agent to control the redox condition in molten fluoride salts.

  6. Behaviour of palladium(II), platinum(IV), and rhodium(III) in artificial and natural waters: Influence of reactor surface and geochemistry on metal recovery

    Energy Technology Data Exchange (ETDEWEB)

    Cobelo-Garcia, Antonio [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)]. E-mail: antonio.cobelo-garcia@plymouth.ac.uk; Turner, Andrew [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Millward, Geoffrey E. [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Couceiro, Fay [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)

    2007-03-07

    The recovery of dissolved platinum group elements (PGE: Pd(II), Pt(IV) and Rh(III)) added to Milli-Q[reg] water, artificial freshwater and seawater and filtered natural waters has been studied, as a function of pH and PGE concentration, in containers of varying synthetic composition. The least adsorptive and/or precipitative loss was obtained for borosilicate glass under most of the conditions employed, whereas the greatest loss was obtained for low-density polyethylene. Of the polymeric materials tested, the adsorptive and/or precipitative loss of PGE was lowest for fluorinated ethylene propylene (Teflon[reg]). The loss of Pd(II) in freshwater was significant due to its affinity for surface adsorption and its relatively low solubility. The presence of natural dissolved organic matter increases the recovery of Pd(II) but enhances the loss of Pt(IV). The loss of Rh(III) in seawater was significant and was mainly due to precipitation, whereas Pd(II) recovery was enhanced, compared to freshwater, because of its complexation with chloride. The results have important implications regarding protocols employed for sample preservation and controlled laboratory experiments used in the study of the speciation and biogeochemical behaviour of PGE.

  7. Fractionated BNCT for locally recurrent head and neck cancer: Experience from a phase I/II clinical trial at Tsing Hua Open-Pool Reactor

    International Nuclear Information System (INIS)

    Wang, Ling-Wei; Chen, Yi-Wei; Ho, Ching-Yin; Hsueh Liu, Yen-Wan; Chou, Fong-In; Liu, Yuan-Hao; Liu, Hong-Ming; Peir, Jinn-Jer; Jiang, Shiang-Huei; Chang, Chi-Wei; Liu, Ching-Sheng; Wang, Shyh-Jen; Chu, Pen-Yuan; Yen, Sang-Hue

    2014-01-01

    To introduce our experience of treating locally and regionally recurrent head and neck cancer patients with BNCT at Tsing Hua Open-Pool Reactor in Taiwan, 12 patients (M/F=10/2, median age 55.5 Y/O) were enrolled and 11 received two fractions of treatment. Fractionated BNCT at 30-day interval with adaptive planning according to changed T/N ratios was feasible, effective and safe for selected recurrent head and neck cancer in this trial. - Highlights: • We treated 12 patients with recurrent Head and Neck (H and N) cancer after radical surgery and radiotherapy since 2010. • Four complete response (CR) and 3 partial response (PR) were found. Total response rate was 58%. • Two patients had local control longer than one year. • No grade 4 or higher toxicity was noted for both acute and chronic effects

  8. Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 01. Projektovanje zastitne komore u hali reaktora RA za rad sa aktivnim eksperimentalnim uredjajima (I-II), II Deo, Album II

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings.

  9. Present status of reactor physics in the United States and Japan-III. 1. Recent Activities of Loading Pattern Optimization Research in Japan

    International Nuclear Information System (INIS)

    Yamamoto, Akio

    2001-01-01

    Because of recent enhancements of optimization algorithms and great improvements in computer hardware, loading pattern (LP) optimization methods are being used as practical design tools both in the pressurized water reactor (PWR) and boiling water reactor (BWR) industries. LP optimization methods are mainly used for the following in-core fuel management activities in Japan: 1. minimization of fuel cycle costs; 2. evaluation of various in-core fuel management scenarios; 3. estimation of the number of feed assemblies needed during several successive cycles for fuel ordering; 4. evaluation of fuel bids. Although engineers can perform these analyses, the major motivations to utilize LP optimization methods are the reduction of manpower and the establishment of engineer-independent LP quality. These are important in today's in-core fuel management tasks. In the following sections, activities related to LP optimization research in Japan are briefly described. The major activity of PWR LP optimization research in Japan is development of the INSIGHT system. The INSIGHT system is an integrated scoping analysis tool for PWRs developed by Nuclear Fuel Industries (NFI). The INSIGHT system is a graphical user interface (GUI)-based interactive design tool that includes LP optimization, automated multicycle analysis, an interactive LP design, core follow, an integrated database, and some auxiliary functions. The INSIGHT system was mainly written in the C++ language and consists of ∼400 000 lines of source code. The GALLOP code is the LP optimization module of the INSIGHT system. An automated multicycle analysis is performed by the MCA code in INSIGHT. The MCA code performs a fuel and burnable poison (BP) inventory search by automatically invoking the GALLOP code, which makes LPs. The MCA code can deal with various constraints that have appeared in practical in-core fuel management, e.g., limitations of fuel/BP stock, forced fuel loading/discharge, limitations of core safety

  10. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  11. Engineering design of ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Wong, C.; Cheng, E.

    1993-07-01

    An efficient organic cooled low activation ferritic steel first wall and shield has been designed for the D- 3 He power reactor ARIES-III. The design allows removal of the large surface heat load without exceeding temperature and stress design limits. The structure is expected to last for the whole reactor life. The major concerns regarding using the organic coolant in fusion reactors have been greatly alleviated

  12. SIMMER-III analytic thermophysical property model

    International Nuclear Information System (INIS)

    Morita, K; Tobita, Y.; Kondo, Sa.; Fischer, E.A.

    1999-05-01

    An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms are designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless steel, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-core materials. (author)

  13. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  14. Considerations Related to LTO for Gen II/III NPP

    International Nuclear Information System (INIS)

    Cojan, Mihail

    2012-01-01

    Today there are some 435 nuclear power reactors operating in 30 countries with a combined capacity of over 372 GWe. In 2011 these provided 2518 billion kWh, about 14% of the world's electricity [1]. The next evolutionary design of Generation III reactors to be deployed over many decades will represent a large part of the worldwide fleet throughout the 21st century. Generation III reactors are the future NPPs with improved safety and reliability, with passive safety systems and with a very low probability for core melt. The objective of this paper is to present the R and D activities that support LTO for Generation II / III Nuclear Reactors. (author)

  15. Richard III

    DEFF Research Database (Denmark)

    Lauridsen, Palle Schantz

    2017-01-01

    Kort analyse af Shakespeares Richard III med fokus på, hvordan denne skurk fremstilles, så tilskuere (og læsere) langt henad vejen kan føle sympati med ham. Med paralleller til Netflix-serien "House of Cards"......Kort analyse af Shakespeares Richard III med fokus på, hvordan denne skurk fremstilles, så tilskuere (og læsere) langt henad vejen kan føle sympati med ham. Med paralleller til Netflix-serien "House of Cards"...

  16. PARDISEKO III

    International Nuclear Information System (INIS)

    Jordan, H.; Sack, C.

    1975-05-01

    This report gives a detailed description of the latest version of the PARDISEKO code, PARDISEKO III, with particular emphasis on the numerical and programming methods employed. The physical model and its relation to nuclear safety as well as a description and the results of confirming experiments are treated in detail in the Karlsruhe Nuclear Research Centre report KFK-1989. (orig.) [de

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  18. Fermilab III

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The total ongoing plans for Fermilab are wrapped up in the Fermilab III scheme, centrepiece of which is the proposal for a new Main Injector. The Laboratory has been awarded a $200,000 Illinois grant which will be used to initiate environmental assessment and engineering design of the Main Injector, while a state review panel recommended that the project should also benefit from $2 million of funding

  19. Fermilab III

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-09-15

    The total ongoing plans for Fermilab are wrapped up in the Fermilab III scheme, centrepiece of which is the proposal for a new Main Injector. The Laboratory has been awarded a $200,000 Illinois grant which will be used to initiate environmental assessment and engineering design of the Main Injector, while a state review panel recommended that the project should also benefit from $2 million of funding.

  20. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

    Long, V.H.; Lam, P.V.; An, T.K.

    1989-01-01

    After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

  1. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  2. Firebird-III program description

    International Nuclear Information System (INIS)

    Lin, M.R.; Prawirosochardjo, S.; Rennick, D.F.; Wessman, E.; Blain, R.J.D.; Wilson, J.M.

    1979-09-01

    The FIREBIRD-III digital computer program is a general network code developed primarily for predicting the thermalhydraulic behaviour of CANDU power reactors during a postulated loss-of-coolant accident and the subsequent emergency coolant injection. Because of its flexibility, the code can also be used to solve a large variety of general two-phase flow problems. This report describes the thermalhydraulic models and the computation methods used in the program

  3. H Reactor

    Data.gov (United States)

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

  4. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  7. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  8. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  9. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  10. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Da Ruan; Benitez-Read, J.S.

    2005-01-01

    Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  12. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  13. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

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

  14. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  15. Issues of high-burnup fuel for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Belac, J.; Milisdoerfer, L.

    2004-12-01

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

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  17. Upgrade of reactor operation technology

    International Nuclear Information System (INIS)

    Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

    2003-01-01

    To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

  18. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  19. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  20. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  1. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  2. Theoretical analysis of nuclear reactors (Phase III), I-V, Part IV, Influence of isotopic composition of nuclear fuel on the reactivity with constant flux; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, IV Deo, Uticaj promene izotopnog sastava goriva na reaktivnost uz konstantan fluks

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Part one of this report presents a series of differential equations describing the nuclear fuel depletion during reactor operation. This series of differential equations is extended to describe the fission products. This part includes equations for effective multiplication factor k{sub eff} and reactivity {rho} as a function of irradiation {tau}. Part two includes results obtained on the analog computer PACE 231 R, and related to Calder Hall type reactor. Part three covers detailed preparation of the series of equations for solution by using the analog computer. Part four includes the list of references related to this task.

  3. Research reactor`s role in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-31

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

  4. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

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

  5. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  7. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  8. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  9. Workshop 95. Part III

    International Nuclear Information System (INIS)

    1995-01-01

    Out of 140 short communications presented in the proceedings, 13 have been inputted into INIS. The topics covered include lifetime control in semiconductor devices by ion irradiation, single crystal scintillation detectors, environmental monitoring, diffusion and sorption of radionuclides in soils, accelerator driven reactors, radioactive waste disposal, digital reactor control systems and research reactors. (Z.S.)

  10. Workshop 95. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    Out of 140 short communications presented in the proceedings, 13 have been inputted into INIS. The topics covered include lifetime control in semiconductor devices by ion irradiation, single crystal scintillation detectors, environmental monitoring, diffusion and sorption of radionuclides in soils, accelerator driven reactors, radioactive waste disposal, digital reactor control systems and research reactors. (Z.S.).

  11. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

  12. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  14. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

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

  15. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  16. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  17. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

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

  18. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  19. Reactor container

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

  1. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  2. Antithrombin III blood test

    Science.gov (United States)

    ... medlineplus.gov/ency/article/003661.htm Antithrombin III blood test To use the sharing features on this page, ... a protein that helps control blood clotting. A blood test can determine the amount of AT III present ...

  3. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

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

  4. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

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

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  8. Reactor container

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  10. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  11. Extraction behaviour of Am(III) and Eu(III) from nitric acid medium in TEHDGA-HDEHP impregnated resins

    Energy Technology Data Exchange (ETDEWEB)

    Saipriya, G.; Kumar, T. [Bhabha Atomic Research Centre Facilities, Kalpakkam (India). Kalpakkam Reprocessing Plant; Kumaresan, R.; Nayak, P.K.; Venkatesan, K.A.; Antony, M.P. [Indira Gandhi Center for Atomic Research, Kalpakkam (India). Fuel Chemistry Div.

    2016-07-01

    The extraction behaviour of Am(III) and Eu(III) from nitric acid medium was studied in the solvent impregnated resins containing extractants such as tetra-bis(2-ethylhexyl)diglycolamide (TEHDGA) or bis-(2-ethylhexyl)phosphoric acid (HDEHP) or mixture of TEHDGA+HDEHP. The rate of extraction of Am(III) and Eu(III) from 1 M nitric acid and the effect of various parameters, such as the concentration of nitric acid in aqueous phase and concentration of TEHDGA and HDEHP in resin phase, on the distribution coefficient of Am(III) and Eu(III) was studied. The distribution coefficient of Am(III) and Eu(III) in HDEHP-impregnated resin decreased and that in TEHDGA-impregnated resin increased, with increase in the concentration of nitric acid. However, in (TEHDGA+HDEHP) - impregnated resin, synergic extraction was observed at lower nitric acid concentration and antagonism at higher nitric acid concentration. The mechanism of Am(III) and Eu(III) extraction in the combined resin was investigated by slope analysis method. The extraction of various metal ions present in the fast reactor simulated high-level liquid waste was studied. The separation factor of Am(III) over Eu(III) was studied using citrate-buffered diethylenetriaminepentaacetic acid (DTPA) solution.

  12. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  13. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  14. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  15. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

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

  16. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

  17. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  18. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  19. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  1. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  2. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  3. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  5. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  6. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  7. Effluent releases at the TRIGA reactor facility

    Energy Technology Data Exchange (ETDEWEB)

    Whittemore, W L [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

  8. The UK MK III GCR experimental physics programme at AEE Winfrith

    Energy Technology Data Exchange (ETDEWEB)

    Johnstone, I

    1972-06-15

    The UK programme of reactor physics experiments in support of the Mk III GCR project started in 1968/69 and has now reached its third main phase. The overall programme is broadly summarised in this report.

  9. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  10. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  11. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  13. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  15. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  16. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  17. Metallothionein (MT)-III

    DEFF Research Database (Denmark)

    Carrasco, J; Giralt, M; Molinero, A

    1999-01-01

    Metallothionein-III is a low molecular weight, heavy-metal binding protein expressed mainly in the central nervous system. First identified as a growth inhibitory factor (GIF) of rat cortical neurons in vitro, it has subsequently been shown to be a member of the metallothionein (MT) gene family...... injected rats. The specificity of the antibody was also demonstrated in immunocytochemical studies by the elimination of the immunostaining by preincubation of the antibody with brain (but not liver) extracts, and by the results obtained in MT-III null mice. The antibody was used to characterize...... the putative differences between the rat brain MT isoforms, namely MT-I+II and MT-III, in the freeze lesion model of brain damage, and for developing an ELISA for MT-III suitable for brain samples. In the normal rat brain, MT-III was mostly present primarily in astrocytes. However, lectin staining indicated...

  18. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

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

  19. SGN III code conversion from Apollo to HP

    International Nuclear Information System (INIS)

    Lee, Hae Cho

    1996-04-01

    SGN III computer code is used to analyze transient behavior of reactor coolant system, pressurizer and steam generators in the event of main steam line break (MSLB), and to calculate mass and energy release for containment design. This report firstly describes detailed work carried out for installation of SFN III on Apollo DN 10000 and code validation results after installation. Secondly, a series of work is also describes in relation to installation of SGN III on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 8 refs. (Author) .new

  20. Application of the reactor kinetics equations to the reactor safety analysis

    International Nuclear Information System (INIS)

    Sdouz, G.

    1976-01-01

    The reactor kinetics equations which can be solved by the computer program AIREK-III are used to describe the behavior of fast reactivity transients. By supplementing this computer program it was possible to solve additional safety problems, e.g. the course of reactor excursions induced by any form of reactivity input, the control of reactivity input as a function of a threshold-energy and the computation of produced energy. (author)

  1. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  7. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  8. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  9. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  12. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  13. 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.)

  14. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  16. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  17. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  18. NNDSS - Table III. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table III. Tuberculosis - 2018.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  19. NNDSS - Table III. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table III. Tuberculosis - 2017.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  20. Workshop 96. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    Part III of the proceedings contain 155 contributions in various fields of science and technology including nuclear engineering, environmental science, and biomedical engineering. Out of these, 10 were selected to be inputted in INIS. (P.A.).

  1. Workshop 96. Part III

    International Nuclear Information System (INIS)

    1995-12-01

    Part III of the proceedings contain 155 contributions in various fields of science and technology including nuclear engineering, environmental science, and biomedical engineering. Out of these, 10 were selected to be inputted in INIS. (P.A.)

  2. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  3. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

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

  4. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

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

  5. Applications of Research Reactors

    International Nuclear Information System (INIS)

    2014-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

  6. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

  7. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  8. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  11. Design improvements in TRIGA reactors

    International Nuclear Information System (INIS)

    Batch, John M.

    1970-01-01

    There have been many design improvements to TRIGA reactor hardware in the past twelve years. One of the more important and most obvious improvements has been in the area of reactor instrumentation. The low profile, completely transistorized Mark III console was a great step forward in a low maintenance, high reliability instrumentation system. Other design improvements include the lazy susan specimen pickup assembly; the specimen container; an empty stainless steel fuel element which can be filled with samples and can be located anywhere in the core; the flexible fuel handling tool; a new fuel measuring tool design; the shock absorber on the adjustable transient rod drive; new testing and evaluation procedures on the thermocouples and other

  12. Regulatory framework and safety requirements for new (gen III) reactors

    International Nuclear Information System (INIS)

    Mourlon, Sophie

    2014-01-01

    Sophie Mourlon, ASN Deputy General Director, described the international process to enhance safety between local safety authorities through organizations such as WENRA. Then she explained to the participants the regulatory issues for the next generation of NPPs

  13. Fusion reactor materials program plan. Section III. Plasma material interaction

    International Nuclear Information System (INIS)

    1978-07-01

    A discussion of materials-related problems and an analysis of such problems is given for each major topical area. The strategy that will be used to solve the materials problems is described. As part of this program strategy, a series of major milestones is identified that extends over the next 20 years. Detailed task descriptions for the next five years leading to the achievement of the major milestones are given. Each task is described on a separate page (or task sheet) which includes the task number, task title, objective, scope, and the major milestones addressed by the task. Secondary milestones within a given task or subtask are defined, together with a priority assignment and an estimate of man-years to accomplish the work. Each Plan is organized along major topics which parallel the Subtask organization of the Task Group responsible for the Plan

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  15. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  16. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  18. Lead reactor strategy economical analysis

    International Nuclear Information System (INIS)

    Ciotti, Marco

    2013-01-01

    Conclusions: • A first attempt to evaluate LFR power plant electricity production cost has been performed; • Electricity price is similar to Gen III + plants; • The estimation accuracy is probably low; • Possible costs reduction could arise from coolant characteristics that may improve safety and simplicity by design; • Accident perception, not acceptable by public opinion, may be changed with low potential energy system (non exploding coolant); • Sustainability improvement could open to a better Public acceptance, depending on us. • Problems may arise in coupling a high capital cost low fuel cost plant in a grid with large amount of intermittent sources with priority dispatch. • Lead fast reactors can compete

  19. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  20. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  1. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  2. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  3. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  4. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  5. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  6. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  8. Fusion Power Demonstration III

    International Nuclear Information System (INIS)

    Lee, J.D.

    1985-07-01

    This is the third in the series of reports covering the Fusion Power Demonstration (FPD) design study. This volume considers the FPD-III configuration that incorporates an octopole end plug. As compared with the quadrupole end-plugged designs of FPD-I and FPD-II, this octopole configuration reduces the number of end cell magnets and shortens the minimum ignition length of the central cell. The end-cell plasma length is also reduced, which in turn reduces the size and cost of the end cell magnets and shielding. As a contiuation in the series of documents covering the FPD, this report does not stand alone as a design description of FPD-III. Design details of FPD-III subsystems that do not differ significantly from those of the FPD-II configuration are not duplicated in this report

  9. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

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

  10. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

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

  11. III-V microelectronics

    CERN Document Server

    Nougier, JP

    1991-01-01

    As is well known, Silicon widely dominates the market of semiconductor devices and circuits, and in particular is well suited for Ultra Large Scale Integration processes. However, a number of III-V compound semiconductor devices and circuits have recently been built, and the contributions in this volume are devoted to those types of materials, which offer a number of interesting properties. Taking into account the great variety of problems encountered and of their mutual correlations when fabricating a circuit or even a device, most of the aspects of III-V microelectronics, from fundamental p

  12. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  13. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  16. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    In the system described the fuel elements are arranged vertically in groups and are supported in such a manner as to tend to tilt them towards the center of the respective group, the fuel elements being urged laterally into abutment with one another. The elements have interlocking bearing pads, whereby lateral movement of adjacent elements is resisted; this improves the stability of the reactor core during refuelling operations. Fuel elements may comprise clusters of parallel fuel pins enclosed in a wrapper of hexagonal cross section, with bearing pads in the form of spline-like ribs located on each side of the wrapper and extending parallel to the longitudinal axis of the fuel element, being interlockable with ribs on pads of adjacent fuel elements. The arrangement is applicable to a reactor core in which fuel elements and control rod guide tubes are arranged in modules each of which comprises a cluster of at least three fuel elements, one of which is rigidly supported whilst the others are resiliently tilted towards the center of the cluster so as to lean on the rigidly supported element. It is also applicable to modules comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements located outside the clusters and also resiliently tilted towards the central voids, the latter being used to accommodate control rod guide tubes. The need for separate structural members to act as leaning posts is thus avoided. Such structural members are liable to irradiation embrittlement, that could lead to core failure. (U.K.)

  18. Summary of Session III

    International Nuclear Information System (INIS)

    Furman, M.A.

    2002-01-01

    This is a summary of the talks presented in Session III ''Simulations of Electron-Cloud Build Up'' of the Mini-Workshop on Electron-Cloud Simulations for Proton and Positron Beams ECLOUD-02, held at CERN, 15-18 April 2002

  19. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  20. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    fuels as well as the applied methodologies. The IRSN proceeds in a relevant and independent assessment of the submitted safety reports. To achieve this goal and maintain over time an independent and relevant assessment capability, the IRSN relies on the excellence of its experts and on state of art techniques and knowledge. The IRSN contributes by its work in key area to cutting edge research and development in order to drive nuclear industry towards making the best use of scientific and technological progress for improving safety, environmental protection and health. To maintain at all times the state of the art knowledge and the operational expertise necessary to deal efficiently with major nuclear accident consequences, the IRSN carries out, on the one hand, its own research and development programs to gain accurate knowledge on still unknown phenomena for safety analysis. On the other hand, the IRSN works out its own scientific calculation methodologies involving industrial calculation chain. Concerning more particularly the 'two-phase flows' thematic, The ISRN must correctly simulate the primary fluid behavior in the reactor in normal operation as well as in accidental situations, to estimate if, in such situations, the core reactor state is fully safe and any safety risk is undergone The research and development programs launched at the ISRN on two-phase flows gather work on advanced thermohydraulic configurations encounter in various reactor states (normal operation or accidental situations), in particular: (i)The estimation of the margin to the critical heat flux in normal operation (DNBR), (ii) The pressurized thermal shock, which is due to mechanical important constraints in the reactor vessel resulting from the injection of a cold fluid in case of emergency cooling (PTS), (iii) The reactivity insertion accident (RIA), (iv) The loss of coolant accident (LOCA), (vi) The accidents in spent-fuel pools and (vii) The severe accident, which could lead to core

  1. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  2. CESEC III code conversion from Apollo to HP9000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hae Cho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new.

  3. CESEC III code conversion from Apollo to HP9000

    International Nuclear Information System (INIS)

    Lee, Hae Cho

    1996-01-01

    CESEC code is a computer program used to analyze transient behaviour of reactor coolant systems of nuclear power plants. CESEC III is an extension of original CESEC code in order to apply wide range of accident analysis including ATWS model. Major parameters during the transients are calculated by CESEC. This report firstly describes detailed work carried out for installation of CESEC III on Apollo DN10000 and code validation results after installation. Secondly, A series of work is also described in relation to installation of CESECIII on HP 9000/700 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

  4. FBR type reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  6. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

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

  7. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  8. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

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

  9. Solar neutrino oscillation parameters after SNO Phase-III and SAGE Part-III

    International Nuclear Information System (INIS)

    Yang Ping; Liu Qiuyu

    2009-01-01

    We analyse the recently published results from solar neutrino experiments SNO Phase-III and SAGE Part-III and show their constraints on solar neutrino oscillation parameters, especially for the mixing angle θ 12 . Through a global analysis using all existing data from SK, SNO, Ga and Cl radiochemical experiments and long base line reactor experiment KamLAND , we obtain the parameters Δm 12 2 =7.684 -0.208 +0.212 x 10 -5 eV 2 , tan 2 θ 12 =0.440 -0.057 +0.059 . We also find that the discrepancy between the KamLAND and solar neutrino results can be reduced by choosing a small non-zero value for the mixing angle θ 13 . (authors)

  10. Modernization of reactor instrumentation for research reactors at Trombay

    International Nuclear Information System (INIS)

    Darbhe, M.D.; Chaudhuri, H.

    1989-01-01

    The three research reactors at Trombay, viz., Apsara, Cirus and Zerlina were commissioned in 1956, 1960 and 1961 respectively. The nuclear instrumentation designs were based on the vacuum tube technology, which was prevalent during those days. The effect of component obsolescence of critical components like vacuum tubes, magnetic amplifiers and sensitrol meter relays was strongly felt since early 1970s. Also, the failure rates of the units were observed to show an increasing trend due to ageing and lack of good quality indigenous spares. Hence it was proposed to replace the nuclear instrumentation units for the three reactors, with those employing modern, state of the art solid state devices, keeping indigenous content as high as practicable. The work started in 1977 with the preparations of specifications and the project was scheduled to be completed in 1981. The project was divided into two phases. The Phase I comprising of nuclear channels common to all reactors and Phase II consisting exclusively of regulating system units of Cirus. The salient stages of project progress and completion were: (i) Fabrication and testing of final design prototypes was completed by end of 1982. (ii) Commissioning of new units at Apsara was completed in January 1984. (iii) Commissioning of new units at Cirus was completed in September 1984. An account of experience in all these stages and problems encountered is given. (author). 6 figs

  11. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

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

  12. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

  13. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

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

  14. Cobalt(III) complex

    Indian Academy of Sciences (India)

    Administrator

    e, 40 µM complex, 10 hrs after dissolution, f, 40 µM complex, after irradiation dose 15 Gy. and H-atoms result in reduction of Co(III) to Co. (II). 6. It is interesting to see in complex containing multiple ligands what is the fate of electron adduct species formed by electron addition. Reduction to. Co(II) and intramolecular transfer ...

  15. Calculus III essentials

    CERN Document Server

    REA, Editors of

    2012-01-01

    REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Calculus III includes vector analysis, real valued functions, partial differentiation, multiple integrations, vector fields, and infinite series.

  16. Thermonuclear reactor

    International Nuclear Information System (INIS)

    Yasutomi, Yoshiyuki; Nakagawa, Moroo; Sawai, Yuichi; Chiba, Akio; Suzuki, Yasutaka.

    1997-01-01

    Silicon composited with reinforcing metals is used for a divertor cooling substrate having an effect as a cooling tube to provide a silicon base composite material having increased electric resistance and toughness. The blending ratio of reinforcing materials in the form of granules, whiskers or long fibers is controlled in order to control heat conductivity, electric resistivity and mechanical performances. The divertor cooling substrate comprising the silicon base composite material is integrated with a plasma facing material. The production method therefor includes ordinary metal matrix composite forming methods such as powder metallurgy, melting penetration method, high pressure solidification casting method, centrifugal casting method and vacuum casting method. Since the cooling plate is constituted with the light metal and highly electric resistant metal base composite material, sharing force due to eddy current can be reduced, and radiation exposure can be minimized. Accordingly, a cooling structure for a thermonuclear reactor effective for the improvement of environmental problems caused by waste disposal can be attained. (N.H.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  19. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

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

  20. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

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

  1. Research reactor's role in Korea

    International Nuclear Information System (INIS)

    Choi, C-O.

    1995-01-01

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960's in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs

  2. ROP design for Enhanced CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

  3. Planning for the decommissioning of a research reactor

    International Nuclear Information System (INIS)

    Dodson, W.J.; Isakari, H.H.; Munro, J.F.; Lim, T.H.; Denton, M.M.; Vernig, P.G.

    1988-01-01

    This paper describes the steps that must be taken and the uncertainties and potential pitfalls that can be encountered in decommissioning a research reactor, whether owned by private industry, a university, or a government agency. The paper is based on the experience in preparing for decommissioning the TRIGA Mark III Berkeley Research Reactor (BRR). Six topics of interest to an owner-operator are addressed: task and schedule planning, decommissioning organization, cost estimating, health and safety considerations, waste management, and regulatory concerns

  4. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

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

  5. EPR by Areva. EPR the 1600+ MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system.

  6. EPR by Areva. EPR the 1600+ MWe reactor

    International Nuclear Information System (INIS)

    2008-01-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

  7. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

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

  8. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

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

  9. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  10. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

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

  11. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  12. 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.)

  13. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

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

  14. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    surge arrester operation during the MSCDN energisation, which causes steep voltage change at the reactor terminal. (ii) Second, the nonuniform voltage distribution, resulting in high stresses across the top inter-turn windings. (iii) Third, the rapid rate-of-change of voltage in the assumed worst-case reactor winding location. This is accompanied by a high dielectric current through the inter-turn winding insulation..

  15. The PISC III programme: 1989 status report

    International Nuclear Information System (INIS)

    Crutzen, S.; Jehenson, P.; McDonald, N.

    1990-01-01

    The PISC Programme has the general objective of assessing procedures and techniques in use for the inspection of pressure components (in particular the vessel and piping). The Series of projects for the Inspection of Steel Components carried out since 1974 under the auspices of the CEC/JRC and the OECD/NEA is a major international effort to better assess the capability and reliability of Non Destructive Inspection procedures on structural components. The programme is now in its third phase (PISC III project); the activities are concentrated on the validation of the PISC II results (e.g. modification of the ASME Inspection Codes) on real structures containing real service defects and the extension of the PISC methodology on most important structural components made of different materials. Most of the PISC test assemblies and structure pieces are representative of (or are coming from) nuclear reactor components. (author)

  16. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  17. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  18. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  19. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  20. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  1. Reactor System Design

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  2. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  3. Nuclear reactors; graphical symbols

    International Nuclear Information System (INIS)

    1987-11-01

    This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

  4. Control for nuclear reactor

    International Nuclear Information System (INIS)

    Ash, E.B.; Bernath, L.; Facha, J.V.

    1980-01-01

    A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

  5. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  6. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2002-01-01

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  7. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  8. continuous stirred tank reactor (CSTR)

    African Journals Online (AJOL)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  9. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Azeez, S.; Alizadeh, A.; Girouard, P.

    2005-01-01

    Full text: The CANDU 6 power reactor is visionary in its approach, remarkable for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Ltd, the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980's as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first CANDU 6 plants- Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea have been in service for more than 21 years and are still producing electricity at peak performance and to the end of 2004, their average lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customer's needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed as the 'Enhanced CANDU 6' (EC6)- which incorporates several attractive but proven features that will make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that will be incorporated in the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  10. Statistical optimization of synthesis procedure and characterization of europium (III) molybdate nano-plates

    Energy Technology Data Exchange (ETDEWEB)

    Pourmortazavi, Seied Mahdi [Malek Ashtar University of Technology, Faculty of Material and Manufacturing Technologies, P. O. Box 16765-3454, Tehran (Iran, Islamic Republic of); Rahimi-Nasrabadi, Mehdi [Imam Hossein University, Nano Science Center, Tehran (Iran, Islamic Republic of); Fazli, Yousef [Islamic Azad University, Department of Chemistry, Faculty of Science, Arak Branch, Arak (Iran, Islamic Republic of); Mohammad-Zadeh, Mohammad [Sabzevar University of Medical Sciences, Department of Physiology and Pharmacology, School of Medicine, Sabzevar (Iran, Islamic Republic of)

    2015-06-15

    Europium (III) molybdate nano-plates were synthesized in this work via chemical precipitation route involving adding of europium (III) ion solution to the aqueous solution of molybdate reagent. Effects of some reaction variables such as concentrations of europium and molybdate ions, flow rate of europium reagent, and reactor temperature on the diameter of the synthesized europium (III) molybdate nano-plates were experimentally investigated by orthogonal array design. The results showed that the size of europium (III) molybdate nano-plates can be optimized by adjusting the concentrations of europium (III) and molybdate ions, as well as the reactional temperature. Europium (III) molybdate nano-plates prepared under the optimum conditions were characterized by X-ray powder diffraction, scanning electron microscopy, and Fourier transform infrared spectroscopy. (orig.)

  11. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

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

  12. Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-09-01

    The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX.

  13. Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

    International Nuclear Information System (INIS)

    1978-09-01

    The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX

  14. Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-11-01

    The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX.

  15. Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

    International Nuclear Information System (INIS)

    1977-11-01

    The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX

  16. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  17. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  18. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  19. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

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

  20. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  1. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    International Nuclear Information System (INIS)

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs

  2. Pseudo Class III malocclusion

    Directory of Open Access Journals (Sweden)

    Fadia M. Al-Hummayani

    2016-04-01

    Full Text Available The treatment of deep anterior crossbite is technically challenging due to the difficulty of placing traditional brackets with fixed appliances. This case report represents a none traditional treatment modality to treat deep anterior crossbite in an adult pseudo class III malocclusion complicated by severely retruded, supraerupted upper and lower incisors. Treatment was carried out in 2 phases. Phase I treatment was performed by removable appliance “modified Hawley appliance with inverted labial bow,” some modifications were carried out to it to suit the presented case. Positive overbite and overjet was accomplished in one month, in this phase with minimal forces exerted on the lower incisors. Whereas, phase II treatment was performed with fixed appliances (braces to align teeth and have proper over bite and overjet and to close posterior open bite, this phase was accomplished within 11 month.

  3. Ammonium diphosphitoindate(III

    Directory of Open Access Journals (Sweden)

    Farida Hamchaoui

    2013-04-01

    Full Text Available The crystal structure of the title compound, NH4[In(HPO32], is built up from InIII cations (site symmetry 3m. adopting an octahedral environment and two different phosphite anions (each with site symmetry 3m. exhibiting a triangular–pyramidal geometry. Each InO6 octahedron shares its six apices with hydrogen phosphite groups. Reciprocally, each HPO3 group shares all its O atoms with three different metal cations, leading to [In(HPO32]− layers which propagate in the ab plane. The ammonium cation likewise has site symmetry 3m.. In the structure, the cations are located between the [In(HPO32]− layers of the host framework. The sheets are held together by hydrogen bonds formed between the NH4+ cations and the O atoms of the framework.

  4. Fast ejendom III

    DEFF Research Database (Denmark)

    Munk-Hansen, Carsten

    Bogen er det tredje bind af tre planlagte bind om fast ejendom: I Overdragelsen, II Bolighandlen og III Ejerbeføjelsen. Fremstillingens giver et grundigt overblik over centrale områder af en omfattende regulering af fast ejendom, med angivelse af litteratur, hvor læseren kan søge yderligere...... oplysning. En ejer af fast ejendom er på særdeles mange områder begrænset i sin råden sammenlignet med ejeren af et formuegode i almindelighed. Fremstillingen tager udgangspunkt i ejerens perspektiv (fremfor samfundets eller myndighedernes). Både den privatretlige og offentligretlige regulering behandles......, eksempelvis ejendomsdannelsen, servitutter, naboretten, hævd, zoneinddelingen, den fysiske planlægning, beskyttelse af natur, beskyttelse af kultur, forurening fra fast ejendom, erstatning for forurening, jordforurening, ekspropriation, byggeri og adgang til fast ejendom....

  5. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

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

  6. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  7. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

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

  8. Rotating reactors : a review

    NARCIS (Netherlands)

    Visscher, F.; Schaaf, van der J.; Nijhuis, T.A.; Schouten, J.C.

    2013-01-01

    This review-perspective paper describes the current state-of-the-art in the field of rotating reactors. The paper has a focus on rotating reactor technology with applications at lab scale, pilot scale and industrial scale. Rotating reactors are classified and discussed according to their geometry:

  9. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  10. Reactor power measuring device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  11. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1981-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drivemechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displayer rods through the reactor vessel

  13. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1982-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drive mechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displacer rods through the reactor vessel. (author)

  14. An overview of future sustainable nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  15. Application of the subchannel analysis code COBRA III C for liquid sodium

    International Nuclear Information System (INIS)

    Nissen, K.L.

    1981-01-01

    The subchannel-analysis code COBRA III C was developed to gain knowledge of mass flow and temperature distribution in rod bundles of light water reactors. A comparison of experimental results for the temperature distribution in a 19 rod bundle with calculations done by the computer program shows the capability of COBRA III C to handle liquid sodium cooling. The code needs sodium properties as well as changed correlations for turbulent mixing and heat transfer at the rod. (orig.) [de

  16. Study on the decommissioning of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Doo Hwan; Jun, Kwan Sik; Choi, Yoon Dong; Lee, Tae Yung; Kwon, Sang Woon; Lee, Jong Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    Currently, KAERI operates TRIGA Mark-II and TRIGA Mark-III research reactors as a general purpose research and training facility. As these are, however, situated at Seoul office site of KAERI which is scheduled to be transferred to KEPCO as well as 30 MW HANARO research reactor which is expected to reach the first criticality in 1995 is under construction at head site of KAERI, decommissioning of TRIGA reactors has become an important topic. The objective of this study is to prepare and present TRIGA facility decontamination and decommissioning plan. Estimation of the radioactive inventory in TRIGA research reactor was carried out by the use of computational method. In addition, summarized in particular were the methodologies associated with decontamination, segmenting processes for activated metallic components, disposition of wastes. Particular consideration in this study was focused available technology applicable to decommissioning of TRIGA research reactor. State-of-the-art summaries of the available technology for decommissioning presented here will serve a useful document for preparations for decommissioning in the future. 6 figs, 41 tabs, 30 refs. (Author).

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  18. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

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

  19. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  20. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  1. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  2. Semiconducting III-V compounds

    CERN Document Server

    Hilsum, C; Henisch, Heinz R

    1961-01-01

    Semiconducting III-V Compounds deals with the properties of III-V compounds as a family of semiconducting crystals and relates these compounds to the monatomic semiconductors silicon and germanium. Emphasis is placed on physical processes that are peculiar to III-V compounds, particularly those that combine boron, aluminum, gallium, and indium with phosphorus, arsenic, and antimony (for example, indium antimonide, indium arsenide, gallium antimonide, and gallium arsenide).Comprised of eight chapters, this book begins with an assessment of the crystal structure and binding of III-V compounds, f

  3. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  4. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

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

  5. Reactor power control device

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  6. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

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

  7. The Maple reactor project

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  8. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

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

  9. Dark matter detection - III

    International Nuclear Information System (INIS)

    Zacek, Viktor

    2015-01-01

    The quest for the missing mass of the universe has become one of the big challenges of todays particle physics and cosmology. Astronomical observations show that only 1% of the matter of the Universe is luminous. Moreover there is now convincing evidence that 85% of all gravitationally observable matter in the Universe is of a new exotic kind, different from the 'ordinary' matter surrounding us. In a series of three lectures we discuss past, recent and future efforts made world- wide to detect and/or decipher the nature of Dark Matter. In Lecture I we review our present knowledge of the Dark Matter content of the Universe and how experimenters search for it's candidates; In Lecture II we discuss so-called 'direct detection' techniques which allow to search for scattering of galactic dark matter particles with detectors in deep-underground laboratories; we discuss the interpretation of experimental results and the challenges posed by different backgrounds; In Lecture III we take a look at the 'indirect detection' of the annihilation of dark matter candidates in astrophysical objects, such as our sun or the center of the Milky Way; In addition we will have a look at efforts to produce Dark Matter particles directly at accelerators and we shall close with a look at alternative nonparticle searches and future prospects. (author)

  10. Safety research for CANDU reactors

    International Nuclear Information System (INIS)

    Hancox, W.T.

    1982-10-01

    Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

  11. Complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III)

    International Nuclear Information System (INIS)

    Ferenc, W.; Bernat, M; Gluchowska, H.W.; Sarzynski, J.

    2010-01-01

    The complexes of 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) have been synthesized as polycrystalline hydrated solids, and characterized by elemental analysis, spectroscopy, magnetic studies and also by X-ray diffraction and thermogravimetric measurements. The analysed complexes have the following colours: violet for Nd(III), white for Gd(III) and cream for Ho(III) compounds. The carboxylate groups bind as bidentate chelating (Ho) or bridging ligands (Nd, Gd). On heating to 1173K in air the complexes decompose in several steps. At first, they dehydrate in one step to form anhydrous salts, that next decompose to the oxides of respective metals. The gaseous products of their thermal decomposition in nitrogen were also determined and the magnetic susceptibilities were measured over the temperature range of 76-303K and the magnetic moments were calculated. The results show that 4-chlorophenoxyacetates of Nd(III), Gd(III) and Ho(III) are high-spin complexes with weak ligand fields. The solubility value in water at 293K for analysed 4-chlorophenoxyacetates is in the order of 10 -4 mol/dm 3 . (author)

  12. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  13. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

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

  14. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Cobian, J.

    2007-01-01

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  15. Measurements at the RA Reactor related to the VISA-2 project - Part 3, calculation of VISA-2 samples influence on the reactor reactivity, on the depression of thermal neutron flux in VK-5 channel and VISA-2 samples; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - III deo, Proracun uticaja uzoraka VISA-2 na reaktivnost reaktora, na depresiju fluksa termalnih neuntrona u kanalu VK-5 i u uzorcima VISA-2

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-01

    The objective of this task was to determine the thermal neutron flux in the RA reactor cell with VISA-2 samples, the influence of VISA-2 samples on the thermal neutron flux distribution in the vicinity of VK-5 channel and the antireactivity of VISA-2 samples placed in the RA reactor core. This distribution is needed for calculating the mean value of absorption and thermal efficiency factor in the reactor cell with the sample. Thermal neutron flux distribution without the sample was calculated by diffusion theory. Fast neutron flux space distribution was assumed to be constant, i.e. that it is dependent only on the nuclear properties of the materials. Macroscopic thermal neutron flux distribution was determined by two-group diffusion theory using the four factor formula applied for two regions: active RA region, and reactor region with calls containing VISA-2 samples. This calculation enables estimation of the VISA-2 samples influence on the thermal neutron flux distribution in the VK-5 channel. Antireactivity of VISA-2 samples was calculated by neutron diffusion theory applying the perturbation method. [Serbo-Croat] Cilj ovog zadatka je da se odredi raspodela fluksa termalnih neutrona u celiji reaktora RA sa uzorcima VISA-2, uticaj uzoraka VISA-2 na makroskopsku raspodelu termalnih neutrona u blizini uzoraka, tj. u kanalu VK-5, kao i antiraktivnost uzoraka VISA-2 ubacenih u jezgro RA. Poznavanje ove raspodele potrebno je za izracunavanje srednje apsorpcije i faktora termalnog iskoriscenja u celiji reaktora sa uzorkom. Raspodela fluksa termalnih neutrona u celiji RA bez uzorka izracunata je na bazi difuzione teorije. Fluks brzih neutrona uzet je kao konstanta s obzirom na prostorne koordinate, tj. da zavisi samo od nuklearnih karakteristika materijala. Makroskopska raspodela fluksa termalnih neutrona u okolini VISA-2 odredjena je na bazi dvogrupne difuzione teorije preko formule cetiri faktora, primenjene na dve zone: prva aktivna zona RA, i druga reaktorka sredina

  16. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Shamsul Amri Sulaiman

    2011-01-01

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

  17. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

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

  18. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  19. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

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

  20. SIMMER-III code-verification. Phase 1

    International Nuclear Information System (INIS)

    Maschek, W.

    1996-05-01

    SIMMER-III is a computer code to investigate core disruptive accidents in liquid metal fast reactors but should also be used to investigate safety related problems in other types of advanced reactors. The code is developed by PNC with cooperation of the European partners FZK, CEA and AEA-T. SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-, time-, and energy-dependent neutron dynamics model. In order to model complex flow situations in a postulated disrupting core, mass and energy conservation equations are solved for 27 density components and 16 energy components, respectively. Three velocity fields (two liquid and one vapor) are modeled to simulate the relative motion of different fluid components. An additional static field takes into account the structures available in a reactor (pins, hexans, vessel structures, internal structures etc.). The neutronics is based on the discrete ordinate method (S N method) coupled into a quasistatic dynamic model. The code assessment and verification of the fluid dynamic/thermohydraulic parts of the code is performed in several steps in a joint effort of all partners. The results of the FZK contributions to the first assessment and verification phase is reported. (orig.) [de

  1. PREFACE: Quantum Optics III

    Science.gov (United States)

    Orszag, M.; Retamal, J. C.; Saavedra, C.; Wallentowitz, S.

    2007-06-01

    All the 50 years of conscious pondering did not bring me nearer to an answer to the question `what is light quanta?'. Nowadays, every rascal believes, he knows it, however, he is mistaken. (A Einstein, 1951 in a letter to M Besso) Quantum optics has played a key role in physics in the last several decades. On the other hand, in these early decades of the information age, the flow of information is becoming more and more central to our daily life. Thus, the related fields of quantum information theory as well as Bose-Einstein condensation have acquired tremendous importance in the last couple of decades. In Quantum Optics III, a fusion of these fields appears in a natural way. Quantum Optics III was held in Pucón, Chile, in 27-30 of November, 2006. This beautiful location in the south of Chile is near the lake Villarrica and below the snow covered volcano of the same name. This fantastic environment contributed to a relaxed atmosphere, suitable for informal discussion and for the students to have a chance to meet the key figures in the field. The previous Quantum Optics conferences took place in Santiago, Chile (Quantum Optics I, 2000) and Cozumel, Mexico (Quantum Optics II, 2004). About 115 participants from 19 countries attended and participated in the meeting to discuss a wide variety of topics such as quantum-information processing, experiments related to non-linear optics and squeezing, various aspects of entanglement including its sudden death, correlated twin-photon experiments, light storage, decoherence-free subspaces, Bose-Einstein condensation, discrete Wigner functions and many more. There was a strong Latin-American participation from Argentina, Brazil, Chile, Colombia, Peru, Uruguay, Venezuela and Mexico, as well as from Europe, USA, China, and Australia. New experimental and theoretical results were presented at the conference. In Latin-America a quiet revolution has taken place in the last twenty years. Several groups working in quantum optics and

  2. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

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

  3. Kuosheng Mark III containment analyses using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Ansheng, E-mail: samuellin1999@iner.gov.tw; Chen, Yen-Shu; Yuann, Yng-Ruey

    2013-10-15

    Highlights: • The Kuosheng Mark III containment model is established using GOTHIC. • Containment pressure and temperature responses due to LOCA are presented. • The calculated results are all below the design values and compared with the FSAR results. • The calculated results can be served as an analysis reference for an SPU project in the future. -- Abstract: Kuosheng nuclear power plant in Taiwan is a twin-unit BWR/6 plant, and both units utilize the Mark III containment. Currently, the plant is performing a stretch power uprate (SPU) project to increase the core thermal power to 103.7% OLTP (original licensed thermal power). However, the containment response in the Kuosheng Final Safety Analysis Report (FSAR) was completed more than twenty-five years ago. The purpose of this study is to establish a Kuosheng Mark III containment model using the containment program GOTHIC. The containment pressure and temperature responses under the design-basis accidents, which are the main steam line break (MSLB) and the recirculation line break (RCLB) accidents, are investigated. Short-term and long-term analyses are presented in this study. The short-term analysis is to calculate the drywell peak pressure and temperature which happen in the early stage of the LOCAs. The long-term analysis is to calculate the peak pressure and temperature of the reactor building space. In the short-term analysis, the calculated peak drywell to wetwell differential pressure is 140.6 kPa for the MSLB, which is below than the design value of 189.6 kPa. The calculated peak drywell temperature is 158 °C, which is still below the design value of 165.6 °C. In addition, in the long-term analysis, the calculated peak containment pressure is 47 kPa G, which is below the design value of 103.4 kPa G. The calculated peak values of containment temperatures are 74.7 °C, which is lower than the design value of 93.3 °C. Therefore, the Kuosheng Mark III containment can maintain the integrity after

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

    International Nuclear Information System (INIS)

    Hirota, Jitsuya

    1984-09-01

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

  5. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  6. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  7. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  8. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

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

  9. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  10. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

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

  11. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

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

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

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

  13. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

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

  14. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

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

  15. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  16. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  17. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  18. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  19. 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.

  20. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

  1. The replacement research reactor

    International Nuclear Information System (INIS)

    Cameron, R.

    1999-01-01

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  2. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  3. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics

  4. 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

  5. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  6. Liquid plugs bouncing against a solid basis, comparison of SIMMER-III and PLEXUS results

    International Nuclear Information System (INIS)

    Maschek, W.; Arnecke, G.; Flad, M.

    1995-01-01

    The SIMMER III code is being tested by application to problems of reactor accidents. The fluid dynamics/thermohydraulics part of the code can be applied also to evaporation/condensation, melt, and freezing phenomena. For a liquid plug bouncing against a solid basis, the momentum transfer is calculated. PLEXUS results turn out to be in significant disagreement with the SIMMER calculations. (orig.)

  7. Základy fúzní energetiky III. – Reaktorové technologie

    Czech Academy of Sciences Publication Activity Database

    Entler, Slavomír; Mlynář, Jan; Dostál, V.

    Srpen (2016), č. článku 14587. ISSN 1801-4399 Institutional support: RVO:61389021 Keywords : Nuclear fusion * reactor technologies * plasma Subject RIV: JF - Nuclear Energetics http://energetika.tzb-info.cz/elektroenergetika/14587-zaklady-fuzni-energetiky-iii-reaktorove-technologie

  8. ATWS: a reappraisal. Part III. Frequency of anticipated transients. Interim report

    International Nuclear Information System (INIS)

    Leverenz, F.L. Jr.; Koren, J.M.; Erdmann, R.C.; Lellouche, G.S.

    1978-07-01

    The document is Part III of the Institute study of the ATWS question. The frequencies of the various events which have led to a reactor scram are documented from the nuclear power plant records. Some of these events, in the absence of scram, could lead to undesirable system response and are the ''transients of significance'' which comprise the anticipated transients of the ATWS question

  9. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  10. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

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

  11. Australian research reactor studies

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  12. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  13. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

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

  14. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-21

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

  15. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  16. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  17. New reactor concepts

    International Nuclear Information System (INIS)

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

    1998-11-01

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

  18. Advanced CANDU reactors

    International Nuclear Information System (INIS)

    Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

    1988-12-01

    AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

  19. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Doshi, P.K.; George, R.A.; Dollard, W.J.

    1982-01-01

    A mechanical spectral shift arrangement for controlling a nuclear reactor includes a plurality of reactor coolant displacer members which are inserted into a reactor core at the beginning of the core life to reduce the volume of reactor coolant-moderator in the core at start-up. However, as the reactivity of the core declines with fuel depletion, selected displacer members are withdrawn from the core at selected time intervals to increase core moderation at a time when fuel reactivity is declining. (author)

  20. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)