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Sample records for acrr reactor

  1. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Kaiser Krista; Chantel Nowlen K.; Russell DePriest K.

    2016-01-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were char...

  2. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  3. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  4. Lessons Learned from Sandia National Laboratories' Operational Readiness Review of the Annular Core Research Reactor (ACRR)

    The Sandia ACRR (a Hazard Category 2 Nuclear Reactor Facility) was defueled in June 1997 to modify the reactor core and control system to produce medical radioisotopes for the Department of Energy (DOE) Isotope Production Program. The DOE determined that an Operational Readiness Review (ORR) was required to confirm readiness to begin operations within the revised safety basis. This paper addresses the ORR Process, lessons learned from the Sandia and DOE ORRS of the ACRR, and the use of the ORR to confirm authorization basis implementation

  5. Design and fabrication of the instrumented fuel elements for the Annular Core Research Reactor (ACRR)

    This report describes the design and fabrication techniques for the instrumented fuel elements of the Annular Core Research Reactor (ACRR). The thermocouple assemblies were designed and fabricated at Sandia Laboratories while the instrumented elements were assembled at Los Alamos Scientific Laboratory. In order to satisfy the ACRR's Technical Specifications, the thermocouples are required to measure temperature in excess of 18000C under rapid heating conditions. Because of the potentially high failure rates for thermocouples in such environments, the instrumented fuel elements are designed so that the thermocouples can be replaced easily

  6. ACRR [Annular Core Research Reactor] fission product release tests: ST-1 and ST-2

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACER) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model. 8 refs., 8 figs., 2 tabs

  7. Neutron Reference Benchmark Field Specification: ACRR Free-Field Environment (ACRR-FF-CC-32-CL).

    Vega, Richard Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity free-field reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.

  8. Radiation Characterization Summary: ACRR Central Cavity Free-Field Environment with the 32-Inch Pedestal at the Core Centerline (ACRR-FF-CC-32-cl).

    Vega, Richard Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Naranjo, Gerald E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lippert, Lance L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    This document presents the facilit y - recommended characteri zation o f the neutron, prompt gamma - ray, and delayed gamma - ray radiation fields in the Annular Core Research Reactor ( ACRR ) for the cen tral cavity free - field environment with the 32 - inch pedestal at the core centerline. The designation for this environmen t is ACRR - FF - CC - 32 - cl. The neutron, prompt gamma - ray , and delayed gamma - ray energy spectra , uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma - ray fluence profiles within the experiment area of the cavity . Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples . Acknowledgements The authors wish to th ank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work . Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  9. Neutron Reference Benchmark Field Specifications: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Environment (ACRR-PLG-CC-32-CL).

    Vega, Richard Manuel [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parm, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integral dosimetry measurements in the neutron field are reported.

  10. Description and characterization of the ACRR's programmable transient rod withdrawal mode

    To satisfy experiment needs for Sandia's Advanced Reactor Safety Program, a programmable Transient Rod Withdrawal (TRW) mode has been developed for the Annular Core Research Reactor (ACRR). The programmable mode is a modification of the existing continuous-withdrawal TRW mode and permits speed and direction changes during the pulse sequence. Basically, a TRW operation is similar to a routine pulse operation except that transient rods are mechanically withdrawn rather than pneumatically fired. Being a pulse-type operation, the TRW mode complies with pulse-mode safety system settings. Control system interlocks prevent the pneumatic firing of rods in the TRW mode. The hardware for the programmable TRW mode includes three ACRR transient rods, the ACRR timer, two rod programmers, a minicomputer and a summing circuit for position indication. Each ACRR transient rod is mechanically driven by a stepping motor (rated torque at 4.24 joules) and is capable of a maximum TRW speed of 26.7 centimeters/ second. The maximum reactivity insertion rate is $2.45/second with a transient rod bank worth of $3.00 and $3.47/second with a bank worth of $4.25, which is expected to be installed soon. The ACRR timer is a multifunctional timer used in all operating modes of the reactor. In the programmable TRW mode, the timer starts the rod programmers and drops regulating rods to terminate the operation. Programmed withdrawal capability is provided by one of two rod programmers (a hardwire-based unit and a microprocessor-based unit). The hardwire unit has eight intervals in which speed, direction and distance are selected by switches on the front panel. The microprocessor-based unit has the capability of 64 intervals in which speed, direction, and distance or time can be specified. Programming this unit is accomplished from the front panel or by inputting data from an HP-9845. minicomputer via a digital I/O interface. Self-test programs in the software provide a continual check of an operating

  11. ACRR fission product release tests: ST-1 and ST-2

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model

  12. Modeling, Calibration, and Verification of a Fission Chamber for ACRR Experimentersa

    Coburn, Jonathan; Luker, S. Michael; Parma, Edward J.; DePriest, K. Russell

    2016-02-01

    When performing research at a reactor facility, experimenters often need to determine the neutron fluence achieved during an operation. Facilities typically provide guidance in the form of neutron fluence per megajoule (MJ) or through passive dosimetry results. After experiment completion, there is sometimes a delay of several days (or weeks) before the passive dosimetry results are available. In the interim, an experimenter does not have confirmation that the desired irradiation levels were reached. Active dosimetry may provide an estimate of neutron fluxes, but few active detectors are available that have been calibrated to measure neutron fluxes obtained inside the Annular Core Research Reactor (ACRR) central cavity environment. For past experiments at the ACRR, the neutron fluence was calculated by integrating the response of a fission chamber rate detection signal and then normalizing this integral to fluence determined from passive dosimetry. An alternative method of directly measuring neutron flux is desired; the new methodology described provides a complete neutron flux profile after a reactor pulse, utilizing fission chamber physics in combination with a compensating ion chamber to extract and convert a current signal to neutron flux as a function of time. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

  13. Modeling, Calibration, and Verification of a Fission Chamber for ACRR Experimentersa

    Coburn Jonathan

    2016-01-01

    Full Text Available When performing research at a reactor facility, experimenters often need to determine the neutron fluence achieved during an operation. Facilities typically provide guidance in the form of neutron fluence per megajoule (MJ or through passive dosimetry results. After experiment completion, there is sometimes a delay of several days (or weeks before the passive dosimetry results are available. In the interim, an experimenter does not have confirmation that the desired irradiation levels were reached. Active dosimetry may provide an estimate of neutron fluxes, but few active detectors are available that have been calibrated to measure neutron fluxes obtained inside the Annular Core Research Reactor (ACRR central cavity environment. For past experiments at the ACRR, the neutron fluence was calculated by integrating the response of a fission chamber rate detection signal and then normalizing this integral to fluence determined from passive dosimetry. An alternative method of directly measuring neutron flux is desired; the new methodology described provides a complete neutron flux profile after a reactor pulse, utilizing fission chamber physics in combination with a compensating ion chamber to extract and convert a current signal to neutron flux as a function of time.

  14. Radiation characterization summary : ACRR 44-inch lead-boron bucket located in the central cavity on the 32-inch pedestal at the core centerline (ACRR-LB44-CC-32-cl).

    Parma, Edward J.,; Quirk, Thomas J.; Lippert, Lance L.; Griffin, Patrick Joseph; Naranjo, Gerald E.; Luker, Spencer Michael

    2013-04-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the 44-inch-long lead-boron bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-LB44-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra are presented as well as radial and axial neutron and gamma-ray flux profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse and steady-state operations are presented with conversion examples.

  15. Neutron Reference Benchmark Field Specification: ACRR 44 Inch Lead-Boron (LB44) Bucket Environment (ACRR-LB44-CC-32-CL).

    Vega, Richard Manuel [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-07-01

    This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the 44 inch Lead-Boron (LB44) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.

  16. Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

    Parma, Edward J.,; Vehar, David W.; Lippert, Lance L.; Griffin, Patrick J.; Naranjo, Gerald E.; Luker, Spencer M.

    2015-06-01

    This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.

  17. Analytical Solution to the Pneumatic Transient Rod System at ACRR

    Fehr, Brandon Michael [Georgia Inst. of Technology, Atlanta, GA (United States)

    2016-01-08

    The ACRR pulse is pneumatically driven by nitrogen in a system of pipes, valves and hoses up to the connection of the pneumatic system and mechanical linkages of the transient rod (TR). The main components of the TR pneumatic system are the regulator, accumulator, solenoid valve and piston-cylinder assembly. The purpose of this analysis is to analyze the flow of nitrogen through the TR pneumatic system in order to develop a motion profile of the piston during the pulse and be able to predict the pressure distributions inside both the cylinder and accumulators. The predicted pressure distributions will be validated against pressure transducer data, while the motion profile will be compared to proximity switch data. By predicting the motion of the piston, pulse timing will be determined and provided to the engineers/operators for verification. The motion profile will provide an acceleration distribution to be used in Razorback to more accurately predict reactivity insertion into the system.

  18. Op weg naar een cleantech toekomst : beknopt verslag van de lustrumconferentie van ACRRES

    PPO Akkerbouw, Groene Ruimte en Vollegrondsgroente

    2013-01-01

    Verslag van de lustrumconferentie van het Application Centre for Renewable REScources (ACRRES) op 7 maart 2013 te Lelystad. De afbeeldingen uit presentaties en de foto’s in het verslag vormen een selectie uit de presentaties en foto’s die zijn gepresenteerd resp. gemaakt tijdens het symposium: Naar

  19. MCNP/MCNPX model of the annular core research reactor.

    DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr. (.,; .)

    2006-10-01

    Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

  20. Critical heat flux prediction for the annular core research reactor

    This paper reports on best estimate predictions of Critical Heat Flux Ratio (CHFR) obtained to support the upgrade of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories for 2 to 4 MWt. The CHF productions are based on the University of New Mexico's (UNM)-CHF correlations in conjunction with the Global Conditions Hypothesis (GCH). Results indicate that for the range of inlet water temperature of 293 to 333 K, CHFR predictions range from 3.9 to 2.1, which is more than sufficient to support the proposed ACRR upgrade

  1. Evaluation of fiber optic performance in the ACRR

    Four fluorosilica clad, all silica core fibers with polyamide buffers were examined for radiation-induced, transient absorption in the central cavity of the Annular Core Research Reactor. The reactor operated 24 times in the pulse mode, typically yielding gamma doses of 15 krad(Si) and neutron fluences of 1.4 x1014 nts/cm2 thermal and 1.0 x 1015 nts/cm2 (fast). The two low-OH fibers absorbed 90% of the light in the 400 to 500 nm region and 30% in the 700 and 800 nm region. The high-OH fibers absorbed 20% in the 400 to 500 nm region and 50% in the 700 to 800 nm region. Saturation of the transient induced absorption was observed in all the fibers. No systematic measurements were taken of long term induced absorption. However, excessive absorption was not a problem in any fibers, even those that received total gamma doses of 5 Mrad(Si). Scintillation in the 680 to 820 mn band was observed. This report documents the data from these experiments

  2. Evaluation of fiber optic performance in the ACRR

    Bodette, D. E.; Brannon, P. J.

    1993-05-01

    Four fluorosilica clad, all silica core fibers with polyamide buffers were examined for radiation-induced, transient absorption in the central cavity of the Annular Core Research Reactor. The reactor operated 24 times in the pulse mode, typically yielding gamma doses of 15 krad(Si) and neutron fluences of 1.4x10(exp 14) nts/cm(sup 2) thermal and 1.0x10(exp 15) nts/cm(sup 2) (fast). The two low-OH fibers absorbed 90% of the light in the 400 to 500 nm region and 30% in the 700 and 800 nm region. The high-OH fibers absorbed 20% in the 400 to 500 nm region and 50% in the 700 to 800 nm region. Saturation of the transient induced absorption was observed in all the fibers. No systematic measurements were taken of long term induced absorption. However, excessive absorption was not a problem in any fibers, even those that received total gamma doses of 5 Mrad(Si). Scintillation in the 680 to 820 mn band was observed. This report documents the data from these experiments.

  3. Cloning, expression, purification, crystallization and preliminary X-ray diffraction analysis of the regulator AcrR from Escherichia coli

    The transcriptional regulator AcrR from Escherichia coli has been cloned, overexpressed, purified and crystallized and X-ray diffraction data have been collected to a resolution of 2.5 Å. This paper describes the cloning, expression, purification and preliminary X-ray data analysis of the AcrR regulatory protein. The Escherichia coli AcrR is a member of the TetR family of transcriptional regulators. It regulates the expression of the AcrAB multidrug transporter. Recombinant AcrR with a 6×His tag at the C-terminus was expressed in E. coli and purified by metal-affinity chromatography. The protein was crystallized using hanging-drop vapor diffusion. X-ray diffraction data were collected from cryocooled crystals at a synchrotron light source. The best crystal diffracted to 2.5 Å. The space group was determined to be P32, with unit-cell parameters a = b = 46.61, c = 166.16 Å

  4. Loop-to-helix transition in the structure of multidrug regulator AcrR at the entrance of the drug-binding cavity

    Manjasetty, Babu A.; Halavaty, Andrei S.; Luan, Chi-Hao; Osipiuk, Jerzy; Mulligan, Rory; Kwon, Keehwan; Anderson, Wayne F.; Joachimiak, Andrzej

    2016-04-01

    Multidrug transcription regulator AcrR from Salmonella enterica subsp. enterica serovar Typhimurium str. LT2 belongs to the tetracycline repressor family, one of the largest groups of bacterial transcription factors. The crystal structure of dimeric AcrR was determined and refined to 1.56 Å resolution. The tertiary and quaternary structures of AcrR are similar to those of its homologs. The multidrug binding site was identified based on structural alignment with homologous proteins and has a di(hydroxyethyl)ether molecule bound. Residues from helices a4 and a7 shape the entry into this binding site. The structure of AcrR reveals that the extended helical conformation of helix a4 is stabilized by the hydrogen bond between Glu67 (helix a4) and Gln130 (helix a7). Based on the structural comparison with the closest homolog structure, the Escherichia coli AcrR, we propose that this hydrogen bond is responsible for control of the loop-to-helix transition within helix a4. This local conformational switch of helix a4 may be a key step in accessing the multidrug binding site and securing ligands at the binding site. Solution smallmolecule binding studies suggest that AcrR binds ligands with their core chemical structure resembling the tetracyclic ring of cholesterol.

  5. Critical heat flux predictions for the Sandia Annular Core Research Reactor

    This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

  6. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  7. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  8. Operational performance of the three bean salad control algorithm on the ACRR

    Ball, Russell M.; Madaras, John J.; Trowbridge, F. Ray; Talley, Darren G.; Parma, Edward J.

    1991-01-01

    Experimental tests on the Annular Core Research Reactor have confirmed that the ``Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute.

  9. Operational performance of the three bean salad control algorithm on the ACRR

    Experimental tests on the Annular Core Research Reactor have confirmed that the ''Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute

  10. Facility modernization Annular Core Research Reactor

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  11. Demonstration of the reactivity constraint approach on SNL's annual core research reactor

    This paper reports on the initial demonstration of the reactivity constraint approach and its implementing algorithm, the MIT-CSDL Non-Linear Digital Controller, on the annual core research reactor (ACCR) that is operated by the Sandia National Laboratories. This demonstration constituted the first use of reactivity constraints for the closed-loop, digital control of reactor power on a facility other than the Massachusetts Institute of Technology's (MIT's) research reactor (MITR-II). Also, because the ACRR and the MITR-II are of very different design, these trials established the generic nature of the reactivity constraint approach

  12. Updated neutron spectrum characterization of SNL baseline reactor environments

    The neutron spectrum characteristics of the primary reactor environments are defined for use by facility customers and to provide an audit trail in support of current quality assurance initiatives. The neutron and gamma environments in the four primary customer environments at SPR-III and ACRR facilities are characterized in detail. Enough detail is provided on other frequently-used environments to support the definition of the 3-MeV and 1-MeV(Si) fluence provided on the Radiation Metrology Laboratory dosimetry reports

  13. Fluorescence-based thermal shift data on multidrug regulator AcrR from Salmonella enterica subsp. entrica serovar Typhimurium str. LT2.

    Manjasetty, Babu A; Halavaty, Andrei S; Luan, Chi-Hao; Osipiuk, Jerzy; Mulligan, Rory; Kwon, Keehwan; Anderson, Wayne F; Joachimiak, Andrzej

    2016-06-01

    The fluorescence-based thermal shift (FTS) data presented here include Table S1 and Fig. S1, and are supplemental to our original research article describing detailed structural, FTS, and fluorescence polarization analyses of the Salmonella enterica subsp. entrica serovar Typhimurium str. LT2 multidrug transcriptional regulator AcrR (StAcrR) (doi:10.1016/j.jsb.2016.01.008) (Manjasetty et al., 2015 [1]). Table S1 contains chemical formulas, a Chemical Abstracts Service (CAS) Registry Number (CAS no.), FTS rank (a ligand with the highest rank) has the largest difference in the melting temperature (ΔT m), and uses as drug molecules against various pathological conditions of sixteen small-molecule ligands that increase thermal stability of StAcrR. Thermal stability of human enolase 1, a negative control protein, was not affected in the presence of various concentrations of the top six StAcrR binders (Fig. S1). PMID:27054155

  14. Reactors

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  15. Updated neutron spectrum characterization of SNL baseline reactor environments. Volume 1, Characterization

    Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

    1994-04-01

    The neutron spectrum characteristics of the primary reactor environments are defined for use by facility customers and to provide an audit trail in support of current quality assurance initiatives. The neutron and gamma environments in the four primary customer environments at SPR-III and ACRR facilities are characterized in detail. Enough detail is provided on other frequently-used environments to support the definition of the 3-MeV and 1-MeV(Si) fluence provided on the Radiation Metrology Laboratory dosimetry reports.

  16. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  17. Development of advanced strain diagnostic techniques for reactor environments.

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  18. Fuel and core testing plan for a target fueled isotope production reactor

    pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

  19. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  20. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  1. Nuclear Reactors

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  2. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Hehr Brian D.

    2016-01-01

    Full Text Available A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field – a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response.

  3. Characterization of Novel Calorimeters in the Annular Core Research Reactor

    Hehr, Brian D.; Parma, Edward J.; Peters, Curtis D.; Naranjo, Gerald E.; Luker, S. Michael

    2016-02-01

    A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response. This work was supported by the United States Department of Energy under Contract DE-AC04-94AL85000. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy.

  4. N Reactor

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  5. Reactor Physics

    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

    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

    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.

  8. Reactor operation

    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

    Russell, Charles R

    2013-01-01

    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. Research Reactors

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  11. Research reactors

    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)

  12. Reactor physics and reactor computations

    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

  13. Research reactors

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  14. Reactor container

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  15. Reactor building

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  16. Heterogeneous reactors

    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)

  17. Plasma reactor

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  18. Reactor physics

    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

  19. Compact Reactor

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  20. NEUTRONIC REACTOR

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  1. NUCLEAR REACTOR

    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.

  2. Nuclear reactors

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  3. Multifunctional reactors

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  4. NUCLEAR REACTOR

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  5. Nuclear reactor

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  6. Breeder reactors

    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

  7. Research reactors - an overview

    West, C.D.

    1997-03-01

    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.

  8. Reactor utilization

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  9. Reactor Neutrinos

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  10. Nuclear reactors

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  11. Nuclear reactors

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  12. Nuclear reactor

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  13. Reactor container

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  14. Reactor building

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  15. Nuclear reactors

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  16. NEUTRONIC REACTORS

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  17. Nuclear reactor

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  18. Nuclear reactor

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  19. Reactor core of nuclear reactor

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  20. Types of Nuclear Reactors

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  1. Nuclear reactor

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  2. Nuclear research reactors

    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.)

  3. Nuclear reactor physics course for reactor operators

    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

  4. Nuclear reactor

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  5. Hybrid adsorptive membrane reactor

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    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.

  6. Hybrid adsorptive membrane reactor

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    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.

  7. Reactor container

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  8. Survey of research reactors

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  9. Department of reactor technology

    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)

  10. RB reactor noise analysis

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  11. Research Nuclear Reactors

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  12. Reactor Physics Training

    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

  13. Introduction of Nuclear Reactor Engineering

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  14. Safeguarding research reactors

    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

  15. Research nuclear reactors

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  16. Nuclear reactor building

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  17. Reactor Physics Programme

    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. Ship propulsion reactors technology

    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.)

  19. Undergraduate reactor control experiment

    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

  20. Process heat reactors

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  1. Reactor System Design

    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. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  3. The Chernobylsk reactor accident

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  4. Zero energy reactor 'RB'

    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)

  5. High solids fermentation reactor

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  6. Fossil nuclear reactors

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  7. Fusion reactor studies

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  8. Reactor power measuring device

    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.)

  9. Light water reactor safety

    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

  10. Nuclear reactor repairing device

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  11. Fundamentals of reactor chemistry

    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)

  12. Nuclear reactor physics

    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

  13. Generation III+ Reactor Portfolio

    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 CO2 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 EPRTM 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 ATMEA1TM 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 KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM 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

  14. The Maple reactor project

    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)

  15. High temperature reactors

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  16. Spinning fluids reactor

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  17. Reactor Safety: Introduction

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  18. Research reactors in Argentina

    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

  19. Thai research reactor

    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)

  20. Nuclear Reactor Physics

    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.

  1. Reactor containment and reactor safety in the United States

    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.)

  2. TRIGA reactor main systems

    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)

  3. Evaluation of research reactors

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  4. Multipurpose research reactors

    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

  5. The nuclear soliton reactor

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  6. Fast Spectrum Reactors

    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...

  7. Fusion reactor research

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  8. The Integral Fast Reactor

    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

  9. The replacement research reactor

    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

  10. PFBR reactor protection

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  11. Reactor BR2

    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

  12. TRIGA reactor characteristics

    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)

  13. Reactor Safety Analysis

    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

  14. Reactor Engineering Department annual report

    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)

  15. Reactor Engineering Division annual report

    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)

  16. Reactor Engineering Division annual report

    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)

  17. New reactor concepts

    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. Reactor construction steels

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  19. Commercialization of fast reactors

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  20. Mirror fusion reactors

    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

  1. Natural convection type reactor

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  2. INVAP's Research Reactor Designs

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  3. Reactor power control device

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  4. Nuclear reactor internals arrangement

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  5. Fusion Reactor Materials

    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

  6. The research reactors their contribution to the reactors physics

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

  7. One piece reactor removal

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  8. Reactor Materials Research

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  9. The fusion reactor

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

  10. Polymerization Reactor Engineering.

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  11. Reactor Materials Research

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  12. Gas-cooled reactors

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  13. Light water reactor program

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  14. Reactor Safety Analysis

    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 two 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. Main achievements in 1999 are reported

  15. Light water type reactor

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  16. Naval propulsion reactors

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  17. Iris reactor conceptual design

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  18. Research reactor DHRUVA

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  19. Reactor core monitoring method

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  20. Physics of nuclear reactors

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  1. Mirror reactor surface study

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  2. FBR type reactor

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  3. TRIGA research reactors

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  4. Status of French reactors

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  5. Nuclear reactor design

    2014-01-01

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

  6. Compact torsatron reactors

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  7. Multi-purpose reactor

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  8. The CAREM reactor and present currents in reactor design

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

  9. Reactor Engineering Division annual report

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

  10. Reactor Engineering Department annual report

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

  11. Reactor Engineering Division annual report

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

  12. Reactor performance calculations for water reactors

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  13. Fourth Generation Reactor Concepts

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  14. Safety of research reactors

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  15. Reactor Engineering Department annual report

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

  16. Reactor engineering department annual report

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

  17. Slurry reactor design studies

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  18. FBR type reactor

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  19. Reactor water sampling device

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  20. Test reactor technology

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  1. Advanced reactor licensing issues

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  2. Nuclear reactor power monitor

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  3. Reactor Sharing Program

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

  4. Determination of research reactor safety parameters by reactor calculations

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  5. Reactor de plasma

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  6. Integral nuclear reactor

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  7. Microfluidic electrochemical reactors

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  8. Licensed operating reactors

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  9. First Algerian research reactor

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  10. Course on reactor physics

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  11. Fast Breeder Reactor studies

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  12. Nuclear reactor theory

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

  13. PWR type reactor

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  14. Fusion Reactor Materials

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  15. International tokamak reactor

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  16. Joyo experimental reactor tour

    JAEA cooperation in remote monitoring focuses on the Joyo Experimental Reactor at the O'arai Research and Development Center. Joyo performs irradiation of test fuels to support development of the fast reactor cycle in Japan, both in international cooperation and in support of the Monju fast reactor, which is now undergoing reconstruction. The tour included an introduction at the model, a visit to the control room, entry into the containment vessel, and viewing of remote monitoring equipment in the Fresh Fuel Storage and at one of the Spent Fuel Ponds. (author)

  17. Fast Breeder Reactor studies

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  18. nuclear reactor design calculations

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  19. Nuclear reactor internal structures

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  20. Reactor monitoring system

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  1. Research reactor support

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  2. Study of power reactor dynamics by stochastic reactor oscillator method

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  3. Nuclear reactor fuel elements

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  4. Reactor power control device

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  5. Reactor pressure boundary materials

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  6. Reactor BR2. Introduction

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  7. Reactor BR2. Introduction

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  8. Reactor parameter simulation system

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  9. Fusion Reactor Materials

    Decreton, M

    2002-04-01

    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.

  10. New reactor type proposed

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  11. Nuclear reactor fuel assembly

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  12. Ageing of research reactors

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  13. Experience with Kamini reactor

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  14. Dossier: research reactors

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  15. Reactor vessel sealing plug

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  16. Future Reactor Experiments

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  17. Reactor hot spot analysis

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  18. Reactor BR2. Introduction

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  19. Research Reactor Benchmarks

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  20. Nuclear reactor (1960)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  1. Reactor Neutrino Spectra

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  2. Pulsed fusion reactors

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  3. Reactor fueling of BWR type reactors

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  4. Reactor Engineering Division annual report

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  5. Special lecture on nuclear reactor

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  6. The replacement research reactor

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  7. OECD Halden reactor project

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  8. Nuclear reactor fuel elements

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  9. Small reactor return

    Current state of the development of present-day small reactors in different countries is performed. Various designs of low and middle power reactors, among which are CAREM (25 MW, PWR), KLT-40 (40 MW, PWR), MRX (30 MW, PWR), IRIS (50 MW, PWR), SMART (1000 MW, PWR), Modular SBWR (50 MW, BWR), PBMR (120 MW, HTGR), GT-HMR (285 MW, HTGR), are discussed

  10. Reactor lattice transport calculations

    The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions used in reactor lattice analysis are given, and finally two basic methods applied for solution of the transport equations are defined. Several remarks on secondary results from lattice transport calculations are added. (author)

  11. Thermal or epithermal reactor

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  12. Future reactor experiments

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  13. Water cooled nuclear reactor

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  14. Jet-Stirred Reactors

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  15. Generation IV reactors: economics

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  16. Future reactor experiments

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  17. Department of Reactor Technology

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  18. AVR reactor physics

    A process for reactivity control was developed and used for fuelling the AVR reactor core, which is largely based on experimentally determined values. By adding fuel elements with different quantities of heavy metals paired with various experimental requirements, great demands were made of reactivity control. Although only a small range of control was available, this was sufficient to operate the reactor and to shut it down safely in the required power and temperature range. (orig.)

  19. Moon base reactor system

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  20. BWR type nuclear reactor

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  1. Emergency reactor scram system

    The present invention provides an emergency reactor scram system capable of shut down a reactor safely upon occurrence of pump trip by improving a passive scram performance for an FBR-type reactor. Namely, a driving motor and an electric generator are connected to a main pump of a primary system. An AC/DC convertor is connected to the electric generator. A shielding plug is disposed to the upper end opening of a reactor container, a control rod drive mechanism is erected on the shielding plug, and an extension pipe is attached to scram magnets of the control rod drive mechanism. The extension pipe is connected to a control rod. The rotation of the shaft of the pump is used as a direct rotator to provide an integrated-type electric generator. The electric generator is electrically connected with the power source of a scram magnet of the emergency scram system. Accordingly, the control rod of the emergency scram system is automatically and rapidly inserted to the reactor core using the power source of the electric generator upon trip of the main pump thereby enabling to scram the reactor safely. (I.S.)

  2. A modular reactor plant

    This paper describes a new concept in liquid metal reactors that is being developed by General Electric under contract to the Department of Energy. This concept is called the Modular Reactor Plant. While this effort is not expected to have a near-term impact, it is directed toward three principal issues currently affecting nuclear power in the United States. First, plant costs have escalated to the point where the startup of new plants require large electric rate increases. Second, the cost of new plants coming on-line today vary by as much as a factor of three. And, third, nuclear construction times often exceed the utilities prudent planning cycle. This paper describes how General Electric's Modular Reactor Plant addreses these issues through shop fabrication and assembly, rail shipment to the site for rapid installation of nuclear components and inherent reactor protection. In addition, it is expected the modular reactor plant will reduce the current cost of development and demonstration of liquid metal reactors to an affordable level

  3. New fission reactor designs

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  4. OECD Halden reactor project

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  5. Reactor power measuring device

    The device of the present invention efficiently calibrates a fixed type gamma ray thermometer of a reactor power measuring device of a BWR type reactor. Namely, the device of the present invention calculates peripheral fuel rod power distribution by calibrating the reactor power distribution by heat generation amount, the reactor power distribution being obtained by a calculation based on a reactor model for converting the signals of a plurality of the gamma ray thermometers in the reactor core based on a conversion formula. In this case, the conversion formula is a relational formula between the power of a thermocouple of the gamma ray thermometer, gamma ray heat generation amount, thermocouple zero power sensitivity relative to a temperature coefficient. A conversion efficient calculation means makes a calibration heater to generate heat at a predetermined power, and the thermocouple zero power sensitivity and the temperature coefficient are obtained based on the output of the gamma ray thermometer in this case. The calibration means updates to conversion type thermocouple zero power sensitivity and temperature coefficient. A calibration execution means executes the operations described above successively, and when the thermocouple zero power sensitivity and the temperature coefficient are out of an allowable range, the means informs it and eliminates the corresponding gamma ray thermometer from the measuring meters. (I.S.)

  6. Reactor safety engineering

    The concept of the work is such that the basic safety philosophy for nuclear power plants as well as the safety features of both types of light water reactors, pressurized and boiling water reactors, and of the fast breeder reactor are dealt with. With the pressurized and boiling water reactors also variations, due to different supplies are mentioned. The state of development considered is characterized by the results of the American reactor safety study having very much influenced the way of presentation and the validity of the information contained. In the introduction the attentive reader is made familiar with the basic traits of safety engineering, the traditional deterministic way of proceeding being supplemented by a detailed illustration of probabilistic means used in the safety analysis. Added to this are comparative descriptions of the individual safety features, their design and mode of operation. There are, e.g., detailed discussion of the emergency core cooling systems, the power supply systems, the reactor protection system, and the containment. Special chapters are attributed to transients with and without the fast shutdown system working and to loss of coolant. The so-called external events are treated somewhat shortly whereas much space is given to core melting problems. The treatment of important events from the safety point of view, including the section on Harrisburg added for reasons of immediate interest, is limited to phenomenological description. (orig.)

  7. Regulations for RA reactor operation

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

  8. The reactor Cabri

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m3/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  9. REACTOR GROUT THERMAL PROPERTIES

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  10. Materials for nuclear reactors

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  11. Reactor physics and economic aspects of the CANDU reactor system

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  12. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  13. Fast breeder reactor research

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  14. BR2 Reactor: Introduction

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3He, 6Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  15. Reactor coolant cleanup facility

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  16. HTGR type reactor

    A reactor core is disposed at the center of a reactor container, a reflector is disposed on the outer side thereof, a steam generator is disposed further outer side thereof coaxially, and they are constituted as an integrated one container. A gas circulator and control rod drives are protruded at the outer side of the lower portion of the integrated container. Heat insulators are disposed on the inner side of the container wall in the upper portion of the reactor container. Helium gas risen in the reactor core and heated to a high temperature descends in a circular steam generator and undergoes heat exchange with water, and is then pressurized in the gas circulator after the lowering of the temperature, and returned to the inlet of the reactor core from the lower central portion of the container. With such procedures, the helium gas as primary coolants circulates only in the container to improve confinement. The device can be reduced in the size and the cost. (I.N.)

  17. Reactor container spray device

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  18. PROTEUS research reactor

    The PROTEUS zero power reactor at the Paul Scherrer Institute (PSI) in Switzerland achieved first criticality in 1968 and since then has been operated as an experimental tool for reactor physics research on test lattices representative of a wide range of reactor concepts. Reactor design codes and their associated data libraries are validated on the basis of the experimental results obtained. PROTEUS is normally configured as a driven system, in which a subcritical test zone is made critical by the surrounding driver zones. The advantages of driven systems can be summarized as follows: - Smaller amount of test fuel is required; - Large range of test zone conditions (including k∞ < 1 states) can be investigated by changes in the driver loading alone, thus avoiding undesirable perturbations to the test zone which would influence the measurement conditions and thus affect the interpretability of the results; - Necessary reactor control and instrumentation equipment (usually perturbing from the experimental viewpoint) can be located in the outer driver regions, thereby avoiding disturbance of the test lattice

  19. Generalities about nuclear reactors

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  20. China experimental fast reactor

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  1. EBT reactor analysis

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  2. Modern research reactors in the world and RA research reactor

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  3. Sodium-cooled nuclear reactors

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  4. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  5. Methanation assembly using multiple reactors

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  6. Mimic of OSU research reactor

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  7. MINT research reactor safety program

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  8. RB research reactor Safety Report

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  9. Fusion reactor materials

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  10. Chernobyl reactor accident

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  11. International Thermonuclear Experimental Reactor

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  12. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  13. Licensed operating reactors

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  14. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  15. Colliding Beam Fusion Reactors

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  16. The MNSR reactor

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  17. Welding and reactor safety

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  18. Reactor operation experience

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  19. Nuclear Rocket Engine Reactor

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  20. Reactor accidents in perspective

    In each of the three major reactor accidents which have led to significant releases to the environment, and discussed in outline in this note, the reactor has been essentially destroyed - certainly Windscale and Chernobyl reactors will never operate and the cleanup operation for Three Mile Island is currently estimated to have cost in excess of US Pound 500 000 000. In each of the accidents there has not been any fatality off site in the short term and any long-term health detriment is unlikely to be seen in comparison with the natural cancer incidence rate. At Chernobyl, early fatalities did occur amongst those concerned with fighting the incident on site and late effects are to be expected. The assumption of a linear non-threshold risk, and hence no level of zero risk is the main problem in communication with the public, and the author calls for simplification of the presentation of the concepts of radiological protection. (U.K.)

  1. Reactor safety equipments

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  2. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  3. Reactor protection system

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  4. Backfitting swimming pool reactors

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  5. MERCHANT MARINE SHIP REACTOR

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  6. Safety systems of heavy water reactors and small power reactors

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  7. Study of future reactors

    Today, more than 420 large reactors with a gross output of close to 350 GWe supply 20 percent of world electricity needs, accounting for less than 5 percent of primary energy consumption. These figures are not expected to change in the near future, due to suspended reactor construction in many countries. Nevertheless, world energy needs continue to grow: the planet's population already exceeds five billion and is forecast to reach ten billion by the middle of the next century. Most less developed countries have a very low rate of energy consumption and, even though some savings can be made in industrialized countries, it will become increasingly difficult to satisfy needs using fossil fuels only. Furthermore, there has been no recent breakthrough in the energy landscape. The physical feasibility of the other great hope of nuclear energy, fusion, has yet to be proved; once this has been done, it will be necessary to solve technological problems and to assess economic viability. Although it is more ever necessary to pursue fusion programs, there is little likelihood of industrial applications being achieved in the coming decades. Coal and fission are the only ways to produce massive amounts of energy for the next century. Coal must overcome the pollution problems inherent in its use; fission nuclear power has to gain better public acceptance, which is obviously colored by safety and waste concerns. Most existing reactors were commissioned in the 1970s; reactor lifetime is a parameter that has not been clearly established. It will certainly be possible to refurbish some to extend their operation beyond the initial target of 30 or 40 years. But normal advances in technology and safety requirements will make the operation of the oldest reactors increasingly difficult. It becomes necessary to develop new generations of nuclear reactors, both to replace older ones and to revive plant construction in their countries that are not yet equipped or that have halted their

  8. AREVA's nuclear reactors portfolio

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  9. Operating US power reactors

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Dec. 31, 1986, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (October, November, and December 1986) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. They are defined as follows: In addition to the tabular data, this article discusses significant occurrences and developments that affected licensed US power reactors during this reporting period. It includes, but is not limited to, changes in operating status, regulatory actions and decisions, and legal actions involving the status of power reactors. We do not have space here for routine problems of operation and maintenance, but such information is available at the Nuclear Regulatory Commission (NRC) Public Document Room, 1717 H Street, NW, Washington, DC 20555. Some significant operating events are summarized elsewhere in this section in the article ''Selected Safety-Related Events,'' and a report on activities relating to facilities still in the construction process is given in the article ''Status of Power-Reactor Projects Undergoing Licensing Review'' in the last section of each issue of this journal. The reader's attention is also called to the regular feature ''General Administrative Activities,'' which deals with more general aspects of regulatory and legal matters that are not covered elsewhere in the journal

  10. Oscillatory flow chemical reactors

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  11. Nuclear reactor simulator

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  12. Experimental reactor physics

    Neutronic experiments in moderators, subcritical assemblies, critical assemblies, and nuclear reactors are described, as well as the techniques of radiation measurements necessary to perform these experiments. Previously dispersed data from government reports, journal articles, and specialized monographs are codified. Original information drawn from the author's experience is included, especially on the pulsed source technique, spectrum measurements, research reactors, and exponential assemblies. The book provides the essential information for carrying out, analyzing, and understanding the experiments. Theory is kept to a minimum. Emphasis is placed on the physics of the situation, and the importance of estimating error as well as the mean value of a measured quantity

  13. Diagnostics for hybrid reactors

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  14. Perspectives on reactor safety

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  15. Clinch River Breeder Reactor

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  16. Netherlands Interuniversity Reactor Institut

    This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

  17. Reactor Materials Research

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  18. Perspectives on reactor safety

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  19. Reactor neutron dosimetry

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  20. Decay of reactor neutrinos

    Vogel, P.

    1984-01-01

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Gösgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit tc.m. / mν is 1-3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1...

  1. Small mirror fusion reactors

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  2. Reactor Materials Research

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  3. Nuclear reactor constructions

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  4. Reactor PIK construction

    The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

  5. Reactor pressure vessel materials

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  6. Reactor gamma spectrometry: status

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  7. Fusion Reactor Materials

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  8. Space-time reactor kinetics for heterogeneous reactor structure

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

  9. Risk prevention during reactor shutdown

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  10. New fast-reactor approach

    The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel

  11. Reactor Engineering Department annual report

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 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. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  12. Reactor operation environmental information document

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

    1989-12-01

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

  13. Operating reactors licensing actions summary

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  14. High Flux Isotope Reactor (HFIR)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  15. The IR-8 reactor operation

    Ryazantsev, E.P.; Egorenkov, P.M.; Yashin, A.F. [Reactor Technology and Materials Research Inst. of RRC ' KI' , Moscow (Russian Federation)

    1997-07-01

    At the Russian Research Center 'Kurchatov Institute' (RRC 'KI') the IR-8 reactor commissioning was carried out in 1981. The reactor was developed in return for earlier existing at RRC 'KI' of the IRT-M reactor (modernized IRT reactor, constructed in 1957). The IRT-M reactor was used for investigations in nuclear physics, solid state physics, radiation chemistry, biology as well as to produce isotopes. Under developing the IR-8 reactor the IRT biological shielding with beam tubes and its process systems were used. The IR-8 reactor creation was founded on application developed by then new fuel assemblies (FA) of IRT-3M type, having two times as great surface of heat transfer and 1.75 times higher U-235 load than the FA of the IRT-2M type, which were used in IRT-M reactor. (author)

  16. Power calibrations for TRIGA reactors

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  17. Reactor operation safety information document

    1990-01-01

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

  18. Reactor safety in Eastern Europe

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  19. VVER and RBMK reactors

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

  20. Thermal Reactor Safety

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  1. Studies on reactor physics

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  2. Nuclear power reactor physics

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  3. Nuclear Reactors and Technology

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  4. Thermal Reactor Safety

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  5. Nuclear reactor container

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  6. ICF tritium production reactor

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  7. Fusion reactor materials

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  8. Nuclear reactor building

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  9. Nuclear reactor constructions

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  10. Fast reactor programme

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  11. Fusion reactor materials

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  12. Reactors. Nuclear propulsion ships

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

  13. Cermet fuel reactors

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  14. Cermet fuel reactors

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  15. Integral Fast Reactor Program

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

  16. Stabilized Spheromak Fusion Reactors

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  17. SRP reactor safety evolution

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  18. Fusion reactor materials

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  19. Department of reactor technology

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  20. The AP1000 reactor

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  1. Fast reactor programme

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  2. Pressure tube type reactor

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  3. TRIGA reactor characteristics

    The general design, characteristics and parameters of TRIGA reactors and fuel are described. This is a training module with the learning objectives: to understand the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and realize the differences between TRIGA fuels and other more traditional. 10 figs., 6 tabs. (nevyjel)

  4. SNAP Nuclear Space Reactors

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  5. Fast breeder reactor

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  6. The Chernobyl reactor accident

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  7. Cermet fuel reactors

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  8. Reactor power measuring device

    The device of the present invention comprises a γ-thermometer disposed in a BWR type reactor, a first amplifier for amplifying the output thereof, a fission ionization chamber disposed in the reactor separately from the γ-thermometer, a second amplifier for amplifying the output thereof, a differential circuit for differentiating the output signal of the second amplifier and a first adding circuit for adding an output signal of the differential circuit and an output signal of the first amplifier. Alternatively, a γ-ray self-powered neutron detector may be disposed instead of the fission ionization chamber. A second adding circuit is also disposed for adding the output signals of plurality of differentiation circuits and inputting the result to the first adding circuit. A sensitivity controller is disposed upstream of the first adder for controlling the sensitivity of the fission ionization chamber. Then, even if time delay should be caused in the γ-thermometer, output signals with good time response characteristic can be obtained by using signals of LPRM or SPND, and currently changing output of the reactor can be measured accurately to provide an effect on the improvement of the safety and operation controllability of the reactor. (N.H.)

  9. Fast reactors: potential for power

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  10. Reactor physics problems on HCPWR

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  11. Nuclear reactor with control rods

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  12. Operating reactors licensing actions summary

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

  13. Reactor regulating and protection system for a light water reactor

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  14. The reactor Pegase

    The reactor Pegase is designed for testing fuel elements for gas-cooled power reactors. Experience has shown that the classical multi-purpose test reactors are not well adapted to these tests. On another side, the introduction of these test elements into the existing power reactors involves numerous problems, which limits their interest. Pegase, which is designed to satisfy these experimental needs, is composed of a parallelepipedal core of enriched Uranium, moderated and cooled by pressurized water. This core is used as a neutron source for eight autonomous loops, containing the elements to be tested, and situated around the core. The core and the eight loops are immersed in a irradiation pool. The loops are placed on the bottom of the pool so, it is possible to move a loop away from the core, or to remove it from the pool without interfering with the operation of the other loops. The irradiation conditions are adjusted, making the synthesis of the following development works. - Experimental studies on Peggy, a zero power critical facility, mock up of Pegase in operation since 1961: measurements of neutron flux level, radial and axial fly distributions on the experiments. Effect of burnable poisons and of movements of the control rods; adjustment of devices (reflectors, screens etc..) needed for optimum performances. - Experimental work on two prototype autonomous loops, heated electrically to the nominal operating power (in operation since 1961): development of the thermodynamic measurements, thermal balances parameters for control of the operating conditions, natural convection. - Studies on Pegase operating under power; thermodynamic measurements on the core circuits on the independent loop circuits; neutronic measurements, etc... The reactor Pegase went critical on the 4. of April 1963 and reached the nominal power of 30 MW on the 28. of May 1963. (authors)

  15. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  16. Reactor physics activities in Japan

    This report reviews the research activity in reactor physics field in Japan during July, 1992 - July, 1993. The review was performed in the following fields : nuclear data evaluation, calculational method development, fast reactor physics, thermal reactor physics, advanced core design, fusion reactor neutronics, nuclear criticality safety, shielding, incineration of radioactive nuclear wastes, noise analysis and control and national programs. The main references were taken from journals and reports published during this period. The research committee of reactor physics is responsible for the review work. (author)

  17. Advanced reactor experimental facilities

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  18. Reactor Engineering Division annual report

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  19. Methanogenesis in Thermophilic Biogas Reactors

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  20. Reactor simulator development. Workshop material

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  1. Reactor water spontaneous circulation structure in reactor pressure vessel

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

  2. Feasible reactor power cutback logic development for an integral reactor

    Han, Soon-Kyoo [KHNP Co., Ltd., Uljin-gun, Gyeong-buk (Korea, Republic of); Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok [Korea Atomic Energy Research Institute (KAERI), Daedeokdaero, Yuseong, Daejeon (Korea, Republic of)

    2013-07-15

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

  3. Feasible reactor power cutback logic development for an integral reactor

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

  4. Elk River Reactor dismantling

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  5. Selecting reactor operator trainees

    Reactor operator trainee selection tends to be more effective if tailored to a utility's unique needs, and offers the organization a better chance for compliance with Federal regulations than if selection methods are adopted without benefit of local research. The costs of operator training range from $50,000 to $100,000. The test validity relative to a variety of training grades and performance measures is reviewed. Of interest is the degree to which tests differentiate reactor operators with respect to simulator training grades and performance in simulator operation; forms of evaluation which have become fairly standard throughout the power industry. The tests administered to each individual were selected because of their presumed relevance to training grades, and the aptitude measures are intended to assess an individual's potential to benefit from training. Tests, availability, form, the abilities they measure, and the time limit are described. (MCW)

  6. Embattled breeder reactor

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  7. Robot for reactor dismantling

    Purpose: To enable to attain the operation on a cylindrical coordinate system thereby performing dismantling operation exactly and at a high reliability. Constitution: A reactor dismantling robot is suspended by ropes by an elevating device to the inside of reactor shielding walls. The robot has a fixed portion having a plurality of legs abutting against the inner surface of the shieling walls while extending and shrinking radially in the horizontal direction and an arm portion having an operation arm disposed with a shielding wall breaking operation device. The arm portion is disposed with a mechanism for vertically moving the operation arm and a mechanism for forwarding and backwarding the operation arm in the horizontal direction and the arm portion itself is constituted so as to be rotatable around a vertical axis. (Seki, T.)

  8. Compact fusion reactors

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  9. The Pegase reactor loops

    After 4 years operation, experimentation and maintenance of the gas loops built especially for the nuclear fuel testing reactor Pegase, it appears desirable not only to gather together in a single document the essential characteristics and particularities of these devices and of their associated equipment, but also to give the reasons for the technical modifications and the way in which they were carried out; this is done here by the persons themselves who were responsible, day after day, for operating these loops. This essentially practically experience thus complements the careful research and preliminary testing carried out on these loops or on their prototypes. It should be of interest to those who deal with problems concerned with the design or operation of irradiation loops in experimental reactors or of similar equipment. (authors)

  10. Spherical torus fusion reactor

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  11. Decay of reactor neutrinos

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Goesgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit t/sup c.m.//m/sub ν/ is 1--3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1--4 MeV could decay into electron-positron pairs. These pairs were not observed, and from the absence of such a signal we deduce restrictions on the corresponding mixing parameters

  12. Advanced boiling water reactor

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  13. Fast reactor database

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

  14. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  15. Gaseous fuel reactor research

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  16. Neutrino physics at reactors

    Reviewing experiments with neutrinos from reactors seems at first to be a simple task, since there were only few. But almost all of them address fundamental questions in particle physics and are of great relevance. This paper reports on these experiments which made use of some of the most sophisticated techniques available at the time they were designed. In these two respects new proposed experiments re in the tradition of the older ones

  17. Decommissioning of research reactors

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  18. Nuclear reactor constructions

    An improvement in the construction of liquid metal cooled nuclear reactors of the kind in which the fuel assembly is submerged in a pool of coolant contained by a primary vessel housed in a concrete vault, is described. In this modification the roof of the vault carries heat exchangers immersed in the pool of coolant, the lower ends of which are hydraulically damped against oscillation caused by seismic disturbances. (U.K.)

  19. Reactor pressure vessel steels

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  20. Fusion Reactor Materials

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  1. Halden reactor project

    The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

  2. The pressurized water reactor

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  3. Towards nuclear fusion reactors

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  4. BWR type reactor

    In a coolant circulation in BWR type reactors, since the mixed stream of steam fluid undergoes a great resistance, the pressure loss due to the flow rate distribution when the coolants flow from the upper plenum into the stand pipe is increased upon passing stand pipe. Also in the spontaneous recycling reactor, pressure loss is still left upon passing the swirling blade of a gas-liquid separator. In view of the above, a plurality of vertical members each having a lower end opened to a gas-liquid two phase boundary and an upper end directly suspended from a steam dryer to the gas-liquid separator. The liquid droplets from the 2-phase boundary heated in the reactor core and formed into a mixed gas-liquid 2-phase stream is directed in the vertical direction accompanied with the steam. The liquid droplets spontaneously fallen by gravity from greater ones successively and the droplets in the steam abutted against the vertical member are fallen as a liquid membrane. Thus, the gas-liquid separation is conducted, the dry steam is directly flown into the steam dryer, thereby capable of providing a gas-liquid separator having gas-liquid separation performance with lower loss than usual. (N.H.)

  5. High temperature gas reactor

    The present invention provides a reflector block structure of a high temperature gas reactor in which graphite blocks are not failed even a containing cylinder loaded to a fuel exchanger collides against to secured reflectors upon loading and withdrawing fuel constitutional elements. Namely, a protection plate made of a metal material such as stainless steel is covered on the secured reflector blocks disposed to the upper most step among secured graphite reflector blocks constituting the reactor core. In addition, positioning guide grooves are formed on the protection plate for guiding the containing cylinder loaded to the fuel exchanger to the column of the reactor core constitutional elements. With such a constitution, even if the containing cylinder of fuel exchanger is hoisted down and collided against the inner circumferential edge of the secured reflector blocks due to deviation of the position and the direction upon exchange of fuels, the reflector blocks are not failed since the above-mentioned portion is covered with the metal protection plate. In addition, the positioning guide grooves lead the fuel exchanger to a predetermined column correctly. (I.S.)

  6. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  7. Reactor water supplementing facility

    Condensates stored in a main condenser are introduced to a turbine-driven reactor feed water pump by way of a low pressure condensate pump, a condensate cleanup facility, a high pressure condensate pump and a low pressure feed water heater by condensate pipelines. The turbine driven feed water pump introduces feed water by way of a high pressure feed water heater to a reactor pressure vessel (RPV). Further, an auxiliary condensate pipeline having a booster pump and connected at one end to the main condenser is connected to the upstream of a motor-driven reactor feed water pump. Downstream of the turbine-driven feed water pump is connected to the downstream of the electromotive feed water pump. Then, when the condensate pump or a turbine-driven feed water pump should stop and if start of a stand-by pump is failed due to some or other reason, the motor-driven feed water pump and the booster pump are started based on a pump stop signal. With such procedures, coolants are supplied to RPV thereby enabling to ensure coolant level in the RPV. (I.N.)

  8. Serious reactor accidents reconsidered

    The chance is determined for damage of the reactor core and that sequel events will cause excursion of radioactive materials into the environment. The gravity of such an accident is expressed by the source term. It appears that the chance for such an accident varies with the source term. In general it is valid that how larger the source term how smaller the chance is for it and vice versa. The chance for excursion is related to two complexes of events: serious damage (meltdown) of the reactor core, and the escape of the liberated radionuclides into the environment. The results are an order of magnitude consideration of the relation between the extent of the source term and the chance for it. From the spectrum of possible source terms three representative ones have been chosen: a large, a medium and a relative small source term. This choice is in accordance with international considerations. The hearth of this study is the estimation of the chance for occurrence of the three chosen source terms for new light-water reactors. refs.; figs.; tabs

  9. Research Reactors of Ukraine

    Ukraine today operates two nuclear research reactors: WWR-M (total capacity of 10 MW), which is located on the site of the Kyiv Nuclear Research Institute of the National Academy of Sciences of Ukraine, and IR-100 (total capacity of 200 kW), which is located on the site of Sevastopol National University of Nuclear Energy and Industry. Both of them have been in operation since the 1960s. The operation project period of WWR-M for which it is licensed is limited to 31 December 2013. In order to improve safety at WWR-M several modernization projects, development of the reactor vessel and the first loop equipment ageing management programme were conducted. According to the license for operation of IR-100 the operation period of the reactor depends upon results from assessments of critical safety elements such as the tank, control and protection system, cable lines and electrical switchgear. Currently the operation period of this equipment has been justified until 2013. (author)

  10. OECD Halden Reactor Project

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  11. Fusion reactor safety

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  12. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  13. BOILER-SUPERHEATED REACTOR

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  14. Fully passive space reactor

    As part of a design project in the Dept. of Nuclear Engineering at Texas A and M Univ., a study was undertaken to develop a fully passive, highly reliable, space reactor system (FPSR) capable of delivering 100 kW(electric) over a 10-yr lifetime. This report summarizes the conceptual design phase and outlines anticipated work on the FPSR project. The purpose of the FPSR is not to compete with, but rather to complement, the SP-100 baseline reactor. The motivation for the new design is to provide a long-lived, high-level power source well suited for remote, unmanned, deep-space missions. In contrast to power systems that would be deployed for cislunar missions in which the power source could be replaced in case of malfunction, the type of missions envisioned for the FPSR system would preclude power package substitution or repair. As a result, the principal design concept for the FPSR is high reliability. This concept is based on the exploitation of fundamental physical laws to eliminate dependency on moving mechanical parts for reactor control, heat removal, and transport and power conversion

  15. Nuclear reactor container

    Upon reactor accident, hydrogen and oxygen are generated by water-zirconium reaction and radiolysis of water, which are accumulated in the reactor. If the concentration of hydrogen and oxygen exceeds a burning limit, there is a possibility of hydrogen burning to cause a danger of deteriorating the integrity of the reactor container and the equipments therein. The limit for the occurrence of the detonation is determined by a relationship between the scale of a detonation cell and the size of the container, and if the scale is greater than the container, the detonation does not occur. The scale of the cell is determined by a gas combustion rate and, if the combustion reaction is suppressed, detonation does not occur even in a large container. Then, an appropriate diluent is added to increase heat capacity of a gas mixture to thereby suppress the temperature elevation of the gas. Incombustible gases having a great heat capacity are preferred for the diluent, and CO2 is used. As the concentration of the CO2 gas to be added is increased, the detonation cell is made greater. Thus, occurrence of detonation due to combustion of the accumulated hydrogen can be prevented. (N.H.)

  16. Reactor core of FBR type reactor

    Gas sealed assemblies are disposed in rows between reactor core fuel assemblies. The gas sealed assembly incorporates inflowed sodium (coolants) and sealed gas in a gas sealing cylinder and an inner hollow of a wrapper tube. A cylindrical heat generating member is disposed in the gas sealing cylinder. The sealed gas is compressed by a discharging pressure of a pump by way of sodium in the wrapper tube. During normal operation, the liquid level of the coolants is present above than a backwarding flow hole, and the temperature of the coolants is raised by the cylindrical heat generation member to raise the temperature of sodium in the backwarding flow hole. High temperature sodium is mixed with low temperature sodium from a lower flow hole at the lower portion of the backwarding flow hole, and sodium at a leak flow hole becomes sodium at a middle temperature. The temperature of the middle temperature sodium is detected by a thermometer. With such procedures, the liquid level in the gas sealed assembly can be detected and confirmed during normal operation. (I.N.)

  17. Backfitting of the FRG reactors

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U3Si2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  18. Nuclear research reactors in Brazil

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  19. Chernobyl reactor transient simulation study

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  20. Water Cooled FBNR Nuclear Reactor

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

  1. Automatic reactor power control device

    Anticipated transient without scram (ATWS) of a BWR type reactor is judged to generate a signal based on a reactor power signal and a scram actuation demand signal. The ATWS signal and a predetermined water level signal to be generated upon occurrence of ATWS are inputted, and an injection water flow rate signal exhibiting injection water flow rate optimum to reactor flooding and power suppression is outputted. In addition, a reactor pressure setting signal is outputted based on injection performance of a high pressure water injection system or a lower pressure water injection system upon occurrence of ATWS. Further, the reactor pressure setting signal is inputted to calculate opening/closing setting pressure of a main steam relief valve and output an opening setting pressure signal and a closure setting pressure signal for the main steam relief valve. As a result, the reactor power and the reactor water level can be automatically controlled even upon occurrence of ATWS due to failure of insertion of all of the control rods thereby enabling to maintain integrity and safety of the reactor, the reactor pressure vessel and the reactor container. (N.H.)

  2. Calibration of RB reactor power

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

  3. Physical security at research reactors

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  4. Progress activities of reactor utilization in 2000

    Charoen, Sakda [Office of Atomic Energy for Peace, Bangkok (Thailand)

    2003-03-01

    Thai Research Reactor - 1/Modification 1(TRR-1/M1) is a multipurpose research reactor with nominal power of 2 MW. The reactor is a swimming pool type, cooled and moderate with light water, using the LEU-fuel. The reactor has been utilized for radioisotope production, neutron beam experiments and reactor physic experiments. The reactor operation data and reactor utilization in 2000 are presented. (author)

  5. Progress activities of reactor utilization in 2000

    Thai Research Reactor - 1/Modification 1(TRR-1/M1) is a multipurpose research reactor with nominal power of 2 MW. The reactor is a swimming pool type, cooled and moderate with light water, using the LEU-fuel. The reactor has been utilized for radioisotope production, neutron beam experiments and reactor physic experiments. The reactor operation data and reactor utilization in 2000 are presented. (author)

  6. Applications of Research Reactors

    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

  7. Reactor noise analysis of experimental fast reactor 'JOYO'

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

  8. Design guide for category V reactors transient reactors

    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 requirements of Category V reactor structures, components, and systems

  9. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  10. Design guide for category IV reactors: liquid metal reactors

    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 requirements of liquid metal cooled fast reactor (Category IV reactor) structures, components, and systems

  11. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

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

  12. Reactor technology: power conversion systems and reactor operation and maintenance

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  13. RA reactor operation, Annex A

    RA reactor was not operated during 1993, although it has been planned to exchange the complete instrumentation until the end of 1993, and to restart the reactor operation in the first half of 1993. This was not fulfilled because the instrumentation was not delivered until the end of 1993. Since our country is under international sanctions it is not possible to plan when the instrumentation will be delivered and foresee the precise date of reactor start-up. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on

  14. Reactor Monitoring with Neutrino Detectors

    The study of the use of neutrino detectors to monitor nuclear reactors is currently a very active field of research. While neutrino detectors located close to reactors have been used to provide information about the global performance of the reactors, a general improvement of the technique is needed in order to use it in a practical way to monitor the fissile contents of the fuel of the nuclear reactors or the thermal power delivered. I describe the current status of the Angra Neutrino Project, aimed to building a low-mass neutrino detector to monitor the Angra II reactor of the Brazilian nuclear power plant Almirante Alvaro Ramos in order to explore new approaches to reactor monitoring with neutrino detectors.

  15. The research reactor TRIGA Mainz

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  16. The molten salt reactor adventure

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF4-ThF4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

  17. Industrializing the liquid metal reactor

    Commercial acceptance of the liquid metal reactor had its beginning with the Fermi reactor, over two decades ago. The pattern of industrialization since that time is discussed, contrasting domestic and foreign experience. The recent termination of the Clinch River reactor project marks a watershed in the U.S. approach towards commercialization. The increased emphasis on achieving cost competitive designs reflects an awareness that barriers to industrialization are institutional and financial, and not technological

  18. Safety device for nuclear reactor

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container

  19. Advances in reactor safety research

    The Nuclear Safety Project is an important part of the German reactor safety research programme. It works on problems concerning safety and environemental risks of LWR reactors and reprocessing plants and investigates accident consequences. At the 1978 annual meeting, the core behaviour on cooling and reactivity disturbances was discussed, as well as release, retention, and possible radiological effects of radioactive pollutants. Among other subjects, fission product retention in LWR reactors and reprocessing plants were reported on as well as hypothetic core meltdown. (orig.)

  20. Turning points in reactor design

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

  1. Optimal control of nuclear reactors

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

  2. Hydrogen Production in Fusion Reactors

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

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  3. CANDU: The fuel conserving reactor

    Because of their high neutron economy and unique design features, CANDU heavy water moderated reactors are the only established commercial reactors able to use directly low fissile content fuels such as natural uranium or uranium recovered from spent light water reactor fuel (RU). These features also help them to achieve the highest fuel utilization of all commercially available reactors, whether the fuel is based on natural uranium or mixed oxides of plutonium, uranium or thorium. As nuclear capacity growth increases demands on the world's finite uranium resources, AECL envisages near term use in CANDU reactors of a fuel incorporating RU and fuels containing thorium, with either plutonium or low enriched uranium (LEU) as the fissile 'driver' fuel. In the long term, AECL proposes the use of future 'Generation X' CANDU reactors with enhanced neutron economy to achieve a near-Self-Sufficient Equilibrium Thorium (SSET) fuel cycle. This CANDU SSET would have a conversion ratio of unity and be able to produce power indefinitely, with the need for little additional fissile material once equilibrium is reached (the amount of 233U needed in the fresh fuel is the same as is present in the discharged fuel, including processing losses.) This would also enable a CANDU-Fast Breeder Reactor (FBR) synergism that would allow each fuel-generating, though expensive, FBR to supply the initial fissile requirements of several less-expensive, CANDU SSET reactors operating on the thorium cycle. The closer the approach to an SSET that CANDUs can achieve, the higher the ratio of CANDUs to breeders in an economically optimized reactor fleet. CANDU reactors thereby become natural partners of both light water-cooled thermal reactors and fast breeder reactors, in both cases making optimum use of their spent fuel components and enhancing the overall sustainability of nuclear power. (authors)

  4. Turning points in reactor design

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  5. Fast reactor programme in India

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  6. Acceptability of reactors in space

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  7. Spiral-shaped disinfection reactors

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  8. Acceptability of reactors in space

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System

  9. Acceptability of reactors in space

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System

  10. Selecting benchmarks for reactor calculations

    Alhassan, Erwin; Sjöstrand, Henrik; Duan, Junfeng; Helgesson, Petter; Pomp, Stephan; Österlund, Michael; Rochman, Dimitri; Koning, Arjan J.

    2014-01-01

    Criticality, reactor physics, fusion and shielding benchmarks are expected to play important roles in GENIV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used to validate computer codes and test nuclear data libraries. However the selection of these benchmarks are usually done by visual inspection which is dependent on the expertise and the experience of the user and there by resulting in a user...

  11. Fast reactors and nonproliferation

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  12. Fast reactor operating experience

    At the beginning of electricity generation from nuclear power there was the breeder, which fulfilled its duty in a number of smaller test and experimental reactors within national programs. Over the years, some of those reactors have attained impressive availabilities, while others have helped to improve our knowledge by the negative results they contributed. Worldwide a decisive step was taken by the mid- to late sixties in the planning and construction of medium sized demonstration fast breeder power plants (250 to 350 MW). In the Federal Republik of Germany, this step is taken belatedly in building the SNR-300. BN-350 in the USSR, Phenix in France, and PFR in the United Kingdom have now been in operation for some ten years. Over that period, valuable experience has been accumulated in sodium technology. The operating behavior of all components and systems working in sodium is called excellent; the hazards associated with sodium, the fire hazard in particular, thus often seem to be greatly overrated. Leakages have been brought under control. It has always been possible so far to trace them back to systemic faults produced in the welding process. The ability of fast sodium cooled reactors to produce more nuclear fuel than they consume has been demonstrated in Phenix, whose breeding ration has been measured to be 1.16. The first true large breeder, Super Phenix in France, is to be commissioned already in 1985. In building another three breeder power plants the European partners in an association hope to achieve the commercial breakthrough of the breeder line. (orig.)

  13. Helias reactor studies

    The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. The magnetic field has 5 field periods and the main optimization principle is the reduction of the Pfirsch-Schlueter currents and the Shafranov shift, which has been verified by computations with the NEMEC and MFBE-codes. The modular coil system comprises 50 coils, which are constructed using NbTi-superconducting cables. The basic dimensions are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 5 T, maximum field on the coils 10 T. Forces and stresses in the coil system have been investigated with the aid of the ANSYS code, which found maximum stress values of about 650 MPa in the coil casing. Helias configurations with 4 and 3 field periods have been constructed by starting from the 5-period case and by eliminating one or two periods while the shape of the coils is kept nearly invariant. In a first survey blanket concepts, developed for the DEMO tokamak, have been adapted to the Helias geometry, in particular, the solid breeder concept developed by FZK (Karlsruhe) has been extrapolated to the Helias geometry identifying the drawbacks and advantages of this concept. Furthermore, the liquid breeder concept using Li7-Pb83 and water-cooling is an interesting alternative for the Helias reactor. Maintenance of blanket and plasma facing components is possible through the portholes between modular coils. Numerical simulations of the start-up phase of the Helias reactor using the TOTAL-P code have confirmed the zero-dimensional modeling of the fusion plasma with the aid of empirical scaling laws. (author)

  14. Improvement of research reactor sustainability

    The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

  15. Management of research reactor ageing

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

  16. Neutrino Oscillation Studies with Reactors

    Vogel, Petr; Zhang, Chao

    2015-01-01

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  17. Advanced Fission Reactor Program objectives

    The objective of an advanced fission reactor program should be to develop an economically attractive, safe, proliferation-resistant fission reactor. To achieve this objective, an aggressive and broad-based research and development program is needed. Preliminary work at Brookhaven National Laboratory shows that a reasonable goal for a research program would be a reactor combining as many as possible of the following features: (1) initial loading of uranium enriched to less than 15% uranium 235, (2) no handling of fuel for the full 30-year nominal core life, (3) inherent safety ensured by core physics, and (4) utilization of natural uranium at least 5 times as efficiently as light water reactors

  18. Reactor safety - an international task

    The dimensions and the significance of the task of ensuring reactor safety can be defined on the basis of experiences gained from Harrisburg and Chernobyl. The countries that use nuclear energy are tied together to a community by virtue of the risk they share. Therefore the GRS is working in close cooperation with the EC, OECD, IAEO and COMECON. This results in safety examinations of the Greifswald reactor, safety analyses of nuclear reactors in Germany, France and the USA and also considerations on the safety demands to be placed on new reactor concepts. (DG)

  19. Studies on nuclear reactor design

    this thesis presents two studies for safety aspects in nuclear reactor design. the fission process that occurs in the reactor core is the most important process for the harmful effect of produced radiation especially neutrons with different energies and gamma radiations for their strong penetrability . so studying the criticality of the fissile materials in the reactor is one of the most important safety aspects for the reactor design, the attenuation of the neutrons and gammas using suitable shielding materials with suitable thicknesses is the second study that is discussed in this thesis

  20. Integrated modular water reactor: IMR

    The Mitsubishi Heavy Industries, Ltd. Has investigated on a concept on small scale reactor with economical efficiency comparable with large scale one. Aims of development on the integrated modular water reactor (IMR) of a small scale reactor plant concept consist in large construction cost reduction through adoption of technique specific to the small scale reactor and integrated production of plural units and in establishment of high safety target without reality in a large scale reactor to realize reduction of operation and maintenance costs by this reduction to simplification of operation and maintenance. Its concrete developmental targets are to make an integrated reactor with vessel size actually producible and the largest output, to remove feasibility of coolant loss accident (LOCA), to remove an accident with feasibility related to fuel fracture, to remove feasibility of nuclear reactor coolant to leak out from a storage vessel, to secure safety of plant without necessity of human and physical assistances from other plants at all on an accident, to make numbers of operators per unit output equal to those of large scale reactor, and to make working amounts at maintenance per unit output equal to large scale reactor by simplification of apparatus practice of rotation on main apparatus such as SG, and so on. Here were described on design concept and plan to realization. (G.K.)

  1. Gas-cooled nuclear reactor

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines

  2. Operating reactors licensing actions summary

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

  3. 3. Interindustry conference on reactor materials science

    This document contains abstracts on papers presented at the Third Interindustry Conference on Reactor Materials Science (Dimitrovgrad, 27-30 October 1992). The subject scope of the papers is a follows: fuel and fuel elements of power reactors; structural materials of fast breeder reactors and thermonuclear reactors; structural materials of WWER and RBMK type reactors; absorbers and moderators

  4. Iris reactor development

    The development progress of the IRIS (International Reactor Innovative and Secure) nuclear power system is presented. IRIS is currently being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. It is aimed at achieving the four major objectives of the Generation IV nuclear systems, i.e., proliferation resistance, enhanced safety, economic competitiveness and reduced waste. The project first year activities, which are summarized here, were focused on core neutronics, in-vessel configuration, steam generator and containment design, safety approach and economic performance. Details of these studies are provided in parallel papers in these proceedings. (author)

  5. Water cooled nuclear reactor

    In the reactor operating with supercritical pressure and temperature part of the water flowing through the moderator tubes is deflected at the outlet and mixed with a residual partial flow of the coolant fed into the core as well as passed along the fuel rods in opposite direction. By special guiding of the flow downward through the guide tubes of the control rods insertion of the control rods is simplified because of reduced frictional forces. By this means it is also achieved to design less critical the control rod cooling with respect to flow rate control and operating behavior in case of a scram. (orig.)

  6. Reactor hall in Plzen

    Technical drawings are shown and described of the Plzen reactor hall. The hall part containing technological equipment consists of aisles 1 and 2, Hall 3M is the pure assembly hall and Hall 3K is the quenching hall. The steel support structures of the respective hall aisles are described. The parameters are given of the crane pathways used. The procedure for hall assembly is shown. Some structural details are described, e.g., the bracket path suspension and the mounting of the crane pathway in aisle 3K showing a detail of the bearing. (J.B.)

  7. Licensed operating reactors

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page

  8. Risk assessment in reactors

    A survey of risk assessment methodology which is being given a lot of attention in the scientific world is presented in this paper. The maximum credible accident approach used for reactor safety assessment in the past has proved inadequate and a probabilistic approach has now caught the fancy of the scientific world. This method took to evaluating single accidents and making them the basis for decisions. This non-probabilistic approach had obvious drawbacks. Most importantly, it undermined less severe accidents which might be more important due to a high frequency of occurrence. (J.A.)

  9. Watercooled atomic reactor

    A hydraulic method for holding fuel assemblies down is described. The nuclear reactor has the vessel divided into an upper high pressure plenum and a lower low pressure or outlet plenum by means of a seal plate. A piston is located at the seal plate with its upper surface exposed to the high pressure and its lower surface exposed to the lower pressure. A push rod attached to the piston extends downwardly in butting relationship with fuel assemblies to hold them down. (C.F.)

  10. Measurement in nuclear reactors

    A nuclear reactor construction has a flux detector comprising a bundle of fibre optics each having a bead incorporating a substance which scintillates on being struck by neutrons or gamma radiations. The other ends of the fibre optics terminate at an image intensifier. The optical fibres may be of glass made from a mixture of silica, alkaline earth metal oxide, cerous oxide and alkali metal oxide. The beads may be incorporated in a disc forming a detector head, which is in a protective guide tube, through which an inert gas may be passed. (author)

  11. TRIGA reactor operating experience

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  12. Reactor system safety assurance

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  13. Reactor coolant pump flywheel

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  14. BWR type reactor

    An austenite/ferrite stainless steel is used for upstream of a condensate cleanup system and only austenitic stainless steel is used for the downstream. An iron concentration in feedwater is kept lower than 0.1ppb and a volume of a reactor cleanup system is increased, to remove Co before it is deposited on the surface of fuels. With such procedures, any of an ion 60Co concentration and a crud 60Co concentration in coolants can be kept low, thereby enabling to suppress radiation dose rate on the surface of equipments and pipelines. (T.M.)

  15. CANDU nuclear reactor technology

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  16. Shutting down two reactors

    Nuclear power will be phased out of the swedish energy system during the first decades of the next century. Commissioned by the swedish government, the National Energy Administration reports a study on the possibilities for, and consequences of, an earlier shut down (1994-1996) of two of the twelve swedish power reactors. Some of the questions studied are: How much will the electricity price raise ?; How will the electricity consumption be affected ?; What are the alternatives to nuclear power ?; What will the cost be ? and What will the environmental effects be ?. (L.E.)

  17. FUEL ASSAY REACTOR

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  18. The Oklo reactors

    The Oklo reactors comprise up to nine 235-U depleted zones in an uranium ore in the Republic of Gabon in West Africa. The depletion in fissile U-235 has been proved to have caused by nuclear chain reactions. The study of the Oklo phenomenon indicates that very efficient retardation mechanisms may operate in nature - at least under special conditions. A closer study of these processes ought to be made to establish the limitations to their occurrence. The Oklo sandstone formation today would probably be considered unacceptable as a host rock for a repository. (EG)

  19. Nuclear reactor plant

    The plant consists mainly of a steam-raising unit and a steam turbine with high pressure and low pressure stages. There is at least one superheater or intermediate superheater between the steam-raising unit and the low pressure stage. In order to improve the plant efficiency, a high temperature reactor is provided as a source of heat for the superheater or intermediate superheater, which supplies the superheater heat with an energy efficiency of over 60%. This increases the overall net efficiency from 33% to over 36%. (orig.)

  20. Licensed reactor nuclear safety criteria applicable to DOE reactors

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

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

    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

  2. Computerized reactor monitor and control for nuclear reactors

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

  3. Reactor Physics Analysis Models for a CANDU Reactor

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  4. Virtual nuclear reactor for education of nuclear reactor physics

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

  5. Multimegawatt Space Reactor Safety

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

  6. Reactor system on barge

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  7. Tokamak experimental power reactor

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

  8. The EPR reactor NDE

    In May 2006, Electricite de France decided to launch the building of the first EPRR Reactor on the Flamanville site in Normandy. The 'Flamanville 3' EPR unit is the first one to be subjected to the French Ministerial Orders of the 10 November 1999 and of the 13 December 2005 from the design phase. According to these orders, the non destructive examination (NDE) planned for the in service inspection (ISI) and for the pre service inspection (PSI) must be operational with a compulsory formal qualification. The PSI is a complete inspection of the main primary and secondary systems. The PSI's objective is to perform before the first core loading all the NDE planned for the future ISI in the same conditions, in order to have a reliable reference for the detection or for the evaluation of the possible damages during the ISI. The 'Flamanville 3' PSI is planned to start end 2010. The program consists of the development and the qualification of the NDE compatible with this new generation reactor's challenges. The paper is about: - the main EPR's objectives and the technological evolutions, - the main component modifications (which have an impact on the NDE), - the place of ISI in the general safety demonstration, - the main inspection objectives, - the NDE qualification process, - the approach to set up the ISI program, - the ISI program. (authors)

  9. Heavy water reactors physics

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors)

  10. Cascade ICF power reactor

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO2. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  11. Reactor refueling machine simulator

    This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console

  12. Decommissioning of commercial reactor

    Yui, Kohei [Japan Atomic Power Co., Tokyo (Japan)

    1997-02-01

    In the case of nuclear reactors, the diversion is often difficult as they are highly purposive, the disassembling is not easy as they are robust, and attention is required to handle the equipment containing radioactive substances. Decommissioning is defined as all the measures taken from the state that facilities become unused to the state of becoming green field. In Japan, already 40 years have elapsed since the effort for nuclear power was begun, and in this paper, the present state and future subjects of the decommissioning of nuclear power stations are summarized at the opportunity that the stop of commercial operation of Tokai Nuclear Power Station was decided recently. In the Tokai Nuclear Power Station, 166 MWe graphite-moderated, carbon dioxide-cooled reactor called improved Calder Hall type is installed, which started the operation in 1966. The circumstances of the decision to stop its operation are explained. The basic policy of the decommissioning of commercial nuclear power stations has been already published by the Advisory Committee for Energy. The state of the decommissioning in various foreign countries is reported. In Japan, the state of green field was realized in 1996 in the decommissioning of the JPDR in Japan Atomic Energy Research institute, and the decommissioning of the atomic powered ship ``Mutsu`` was completed. (K.I.)

  13. Molten salt reactor type

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF2-ThF4-UF4) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate

  14. Canada's reactor exports

    A brief sketch of the development of Canada's nuclear exports is presented and some of the factors which influence the ability to export reactors have been identified. The potential market for CANDUs is small and will develop slowly. The competition will be tough. There are few good prospects for immediate export orders in the next two or three years. Nonetheless there are reasonable opportunities for CANDU exports, especially in the mid-to-late 1980s. Such sales could be of great benefit to Canada and could do much to sustain the domestic nuclear industry. Apart from its excellent economic and technical performance, the main attraction of the CANDU seems to be the autonomy it confers on purchasing countries, the effectiveness with which the associated technology can be transferred, and the diversification it offers to countries which wish to reduce their dependence on the major industrial suppliers. Each sales opportunity is unique, and marketing strategy will have to be tailored to the customer's needs. Over the next decade, the factors susceptible to Canadian government action which are most likely to influence CANDU exports will be the political commitment of the government to those reactor exports, the performance established by the four 600 MWe CANDUs now nearing completion, the continuing successful operation of the nuclear program in Ontario, and the co-ordination of the different components of Canada's nuclear program (AECL, nuclear industry, utilities, and government) in putting forth a coherent marketing effort and following through with effective project management

  15. Space reactor shield technology

    The reactor shield mass contributes a large portion (10% to 25%) to the total mass of an unmanned reactor system. Different shield materials are required to attenuate neutrons and gamma rays and still obtain a minimum mass. The shield material selection should also consider structural characteristics, physical and chemical properties, fabricability and availability. Minimum mass is achieved by using a shadow shield. Neutron capture gamma ray and heat generation are extremely important considerations. Lithium hydride was selected for the neutron shield material due to its excellent properties. It has to be canned and may be compartmentalized to reduce the probability of complete shielding effectiveness loss due to meteoroid puncture of the can. The initial shield design was based on previous SNAP shield design experience. The Monte Carlo Neutron Photon code, which includes the radiation scattering with the radiator and power conversion system, was then used to ensure that the design requirements were met. Fabrication of the shield by casting techniques is recommended to maintain shield integrity during vibration and to accommodate complex penetrations. A method for casting full-scale shields is described

  16. Innovative hybrid biological reactors using membranes

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  17. 2012 review of French research reactors

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

  18. Control rod drive of nuclear reactor

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  19. Reactor Neutrino Physics -- An Update

    Boehm, Felix

    1999-01-01

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  20. Status of Japanese university reactors

    Fujita, Yoshiaki [Research Reactor Institute, Kyoto Univ., Kumatori, Osaka (Japan)

    1999-08-01

    Status of Japanese university reactors, their role and value in research and education, and the spent fuel problem are presented. Some of the reactors are now faced by severe difficulties in continuing their operation services. The point of measures to solve the difficulties is suggested. (author)

  1. High Pressure Boiling Water Reactor

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  2. United States fast reactor programme

    The fast reactor programs in USA deal with: EBR-II termination program; fast flux test facility; nuclear energy research initiative; accelerator transmutation of waste; and other non-DOE funded activities. These are concerned with preliminary concept development activities of a secure transportable autonomous reactor -liquid metal cooled

  3. NPI's work on future reactors

    Nuclear Power International (NPI) is a partnership between Framatome and Siemens/KWU to market their existing reactors and to manage the common development of a new generation of nuclear reactors and nuclear power plants. The main objectives are discussed. 1 fig

  4. IAEA safeguards at research reactors

    The International Atomic Energy Agency applies safeguards to almost 150 facilities classified as research reactors. From a safeguards point of view, these facilities present a spectrum of features that must be addressed both from the nuclear material and from the operational viewpoints. The nuclear fuel used by these reactors varies from high enriched uranium (NEU), up to 93 U-235, to natural uranium and the thermal power output from over 100 megawatt to less than ten watts. Research reactors are also used for a wide variety of purposes, including materials testing, radiosotope production, training and nuclear physics studies. The effort spent by the Agency in safeguarding these reactors is dependant upon the thermal power of the reactor and on the quantity and type of nuclear material present. On some research reactors, the Agency devotes more inspection effort than on a large power reactor. On others, very little effort is reguired. Safeguards that are applied are done so according to Agency State agreements and consist of a combiination of nuclear material accounting and containment and surveillance. In this paper, the safeguards activities performed by the State and by the Agency will be reviewed for a large (≤50MWt) and for a small (≥ 1 MWt) reactor according to the most common type agreement. (author)

  5. Kyoto University Reactor diagnostic system

    For the safety of a nuclear reactor, it is very important that the operators and manager make exact judgement about the various conditions of the nuclear reactor occurring at times. The research is advanced for the purpose of adopting a computer system for the research reactor of Kyoto University (KUR), offering effective information to operators and maintenance workers, making the advice for exactly judging the conditions of the reactor by sufficiently grasping them, consequently, developing the system for increasing the safety of the reactor. For the development of this system, also technical officials took part positively and cooperated in the research and development based on the experience of the maintenance and operation of the research reactor carried out daily. The system comprises the data acquisition part, data base, abnormality diagnostic part, man-machine interface part, and individual dealing part. The abnormality of the reactor is identified by the judgement of operators by referring to the data memorized in the data base, then, the reactor is operated. The constitution of the computer system used is shown. The CPU is a minicomputer ECLIPSE S-140, and the main memory is 512 kB. The auxiliary memories are a fixed disk equipment of 73 MB, two floppy disk equipments and a magnetic tape equipment. Respective subsystems are explained. (Kako, I.)

  6. Brookhaven leak reactor to close

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  7. Status of Japanese university reactors

    Status of Japanese university reactors, their role and value in research and education, and the spent fuel problem are presented. Some of the reactors are now faced by severe difficulties in continuing their operation services. The point of measures to solve the difficulties is suggested. (author)

  8. Enrichment reduction for research reactors

    The worldwide activities on enrichment reduction for research reactors are reviewed and the national and international programs are described. Especially the following points are discussed: Benchmark calculations, reactor safety, fuel element development, irradiation tests, post irradiation examinations, full core demonstrations, activities of the GKSS and economical questions. (orig.)

  9. Proton Collimators for Fusion Reactors

    Miley, George H.; Momota, Hiromu

    2003-01-01

    Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.

  10. Physics of gas-cooled reactors

    This paper deals with technical aspects of the gas-cooled reactors safety and design. It content five main parts: the new safety requirements for innovative reactors; the concept of a non-meltable nuclear reactor; the new concepts of decay heat removal from the reactor system; the behaviour of reactor system in case of extreme reactivity accidents; technical innovations in the reactor technology and necessary developments and proof tests. (A.L.B.)

  11. Cooling system for reactor container

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  12. Simulator for materials testing reactors

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  13. Replacement research reactor for Australia

    In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

  14. Antineutrino Monitoring of Thorium Reactors

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  15. Directions in advanced reactor technology

    Successful nuclear power plant concepts must simultaneously performance in terms of both safety and economics. To be attractive to both electric utility companies and the public, such plants must produce economical electric energy consistent with a level of safety which is acceptable to both the public and the plant owner. Programs for reactor development worldwide can be classified according to whether the reactor concept pursues improved safety or improved economic performance as the primary objective. When improved safety is the primary goal, safety enters the solution of the design problem as a constraint which restricts the set of allowed solutions. Conversely, when improved economic performance is the primary goal, it is allowed to be pursued only to an extent which is compatible with stringent safety requirements. The three major reactor coolants under consideration for future advanced reactor use are water, helium and sodium. Reactor development programs focuses upon safety and upon economics using each coolant are being pursued worldwide. These programs are discussed

  16. Automated reactor records evaluation framework

    The only truly reliable method for core physics code validation is comparison against experimental data – and for power nuclear reactors, the only reasonably acquirable kind of experimental data are the reactor records. However, the amount of the data coming from the reactor operation is often so vast that it can be discouraging for the code developers to use it properly. Thus, the validation package is further reduced because the data is hard to use. This paper presents an elaborate, fully automated framework, which was designed and implemented in our institute, for reactor records processing and its use for core physics code validation. The workflow, implemented as a Web 2.0 application, provides a practical and painless solution for use of reactor records data for code development and validation. (author)

  17. Fast reactors and nuclear nonproliferation

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  18. Reactor-Produced Medical Radionuclides

    Mirzadeh, S.; Mausner, L. F.; Garland, M. A.

    The therapeutic use of radionuclides in nuclear medicine, oncology, and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by β- emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for the production of a number of medical radionuclides in a reactor.

  19. Engineering reactors for catalytic reactions

    Vivek V Ranade

    2014-03-01

    Catalytic reactions are ubiquitous in chemical and allied industries. A homogeneous or heterogeneous catalyst which provides an alternative route of reaction with lower activation energy and better control on selectivity can make substantial impact on process viability and economics. Extensive studies have been conducted to establish sound basis for design and engineering of reactors for practising such catalytic reactions and for realizing improvements in reactor performance. In this article, application of recent (and not so recent) developments in engineering reactors for catalytic reactions is discussed. Some examples where performance enhancement was realized by catalyst design, appropriate choice of reactor, better injection and dispersion strategies and recent advances in process intensification/ multifunctional reactors are discussed to illustrate the approach.

  20. Description of the RBMK reactors

    The RBMK type reactors are described. The chief design features are vertical pressure tubes with onload refuelling, fuel assemblies embodying 18-pin clusters, each pin comprised of uranium dioxide pellets clad in a zirconium alloy tube, a graphite moderator and reflector and boiling water coolant with the steam going to the turbines. The general specifications and characteristics of the RBMK-1000 fuel assembly and fuel element are tabulated. The reactor is described in detail. Illustrations and circuit diagrams are given at the end of the section. The appendix reviews the performance characteristics of the RBMK reactor which are relevant to an understanding of the Chernobyl accident. Interspersed with the description of the reactor are comments comparing UK reactor types, operating experience and safety regulations as laid down by the Nuclear Installations Inspectorate. These show the aspects of the RBMK design which would not be acceptable in the UK. (U.K.)