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Sample records for brookhaven graphite research reactor

  1. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0

    Energy Technology Data Exchange (ETDEWEB)

    Evan Harpenau

    2011-07-15

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

  2. Burn-Up Calculations for the Brookhaven Graphite Research Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Fuel bum-up calculations for the Brookhaven Graphite Research Reactor involve a distribution of the thermal megawatt days of operations to the fuel elements in proportion to the average thermal neutron flux at their location in the reactor. The megawatt days so assigned can be converted to equivalent uranium-235 consumption when needed. The original fuel loading for the BGRR was neutral uranium and a single calculation was performed on each fuel element upon discharge from the reactor. A subsequent change to a fully enriched uranium-235 fuel element, however, introduced complications. The average loading of enriched uranium involves about 4800 individual elements, each occupying four different reactor positions during its term in the reactor. The total term for a central channel element is about one year as against six to eight years for an element in a peripheral channel. With the large number of individual fuel elements involved and the approximately monthly small changes needed for operation, it was necessary to resort to a computer programme to follow the burn-up of all the elements on the reactor continuously. Both this and other functions of the computer programme are discussed in the paper. To date, uranium has been recovered from two batches of spent fuel. On the first, involving 3674 elements discharged from the reactor over a period of 4.9 years, the recovery figures were 5.5% higher than the calculated total of 32.3 kg uranium-235. On the second batch, involving 1296 elements discharged from the reactor over a period of one year, the recovery figures were 2.3% higher than the calculated figures of 10.8 kg uranium-235. This relatively close agreement seems to indicate that the assumptions made to simplify the programme are acceptable and that the results of the programme are satisfactory for our particular accounting and operating requirements. (author)

  3. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    International Nuclear Information System (INIS)

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at

  4. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    International Nuclear Information System (INIS)

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  5. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    Energy Technology Data Exchange (ETDEWEB)

    Kneitel, Terri [US DOE, Brookhaven Site Office (United States); Rocco, Diane [Brookhaven National Laboratory (United States)

    2012-07-01

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  6. Seismic research on graphite reactor core

    International Nuclear Information System (INIS)

    Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)

  7. Brookhaven leak reactor to close

    CERN Multimedia

    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).

  8. Epithermal beam development at the BMRR [Brookhaven Medical Research Reactor]: Dosimetric evaluation

    International Nuclear Information System (INIS)

    The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al2O3) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs

  9. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR)

    International Nuclear Information System (INIS)

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm3 or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study. (author)

  10. Spectral characterization of the epithermal-neutron beam at the Brookhaven medical research reactor

    International Nuclear Information System (INIS)

    The power burst facility boron neutron capture therapy (PBF/BNCT) program schedule required the use of an epithermal-neutron beam before the PBF would be available. The beam was needed to carry out the acute, dose-tolerance study on healthy canines and the treatment protocol on spontaneous tumor canines. Calculations on available U.S. test reactors confirmed that the Brookhaven medical research reactor (BMRR) would be capable of providing an epithermal-neutron beam with sufficient intensity while limiting the fast-neutron and gamma dose contamination to acceptable levels for the canine irradiation studies. A joint Idaho National Engineering Laboratory (INEL)/Brookhaven National Laboratory (BNL) program was instituted to design, construct, install, and measure the performance of an epithermal-neutron beam filter for the BMRR. Aluminum oxide was selected as the filter material because it provided the desired neutron spectrum characteristics given the physical constraints of the available BMRR irradiation beam port. Neutron spectrum measurements of the exit beam were undertaken by INEL as a means to evaluate the performance of the new filter and the validity of neutron transport calculations. The preliminary data from activation measurements were presented at the Neutron Beam Design Workshop at Massachusetts Institute of Technology (MIT) in March 1989. The updated activation results and the proton-recoil measurements are presented in this paper and are compared with predictions derived from a two-dimensional transport calculation

  11. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source

    International Nuclear Information System (INIS)

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry

  12. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 1010 n/cm2·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10-11 cGy·cm2/nth) and fast neutron (0.58 x 10-11 cGy·cm2/nth) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  13. Neutron spectrum measurements in the aluminum oxide filtered beam facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Neutron spectrum measurements were performed on the aluminum oxide filter installed in the Brookhaven Medical Research Reactor (BMRR). For these measurements, activation foils were irradiated at the exit port of the beam facility. A technique based on dominant resonances in selected activation reactions was used to measure the epithermal neutron spectrum. The fast and intermediate-energy ranges of the neutron spectrum were measured by threshold reactions and 10B-shielded 235U fission reactions. Neutron spectral data were derived from the activation data by two approaches: (1) a short analysis which yields neutron flux values at the energies of the dominant or primary resonances in the epithermal activation reactions and integral flux data for neutrons above corresponding threshold or pseudo-threshold energies, and (2) the longer analysis which utilized all the activation data in a full-spectrum, unfolding process using the FERRET spectrum adjustment code. This paper gives a brief description of the measurement techniques, analysis methods, and the results obtained

  14. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  15. Hypernuclear physics research at Brookhaven

    International Nuclear Information System (INIS)

    This paper describes the results of a recently completed study of the hypernucleus 12/sub Λ/C. The observed formation of hypernuclear states at large momentum is compared with theoretical expectations. Future directions of the research program at Brookhaven are outlined

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

    International Nuclear Information System (INIS)

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

  17. Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

    Science.gov (United States)

    Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

    1994-10-01

    Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed. PMID:7869995

  18. Radiation Dosimetry in the BNCT Patient Treatment Room at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    The BMRR was a 3 MW light water reactor that had an epithermal neutron beam that was used to perform clinical trials on patients with malignant brain tumors. A series of measurements and calculations had been performed in the treatment room both prior to the trials and during the trials. The details of the measurements and the Monte Carlo calculations are presented and compared.

  19. OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR

    International Nuclear Information System (INIS)

    Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use

  20. Distribution of the thermal neutron field around the graphite reflector of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Thermal neutron flux distributions around the graphite reflector of the Dalat Nuclear Research Reactor are determined by the method for neutron activating Cu foils. The major results are as follows: a/The axial distributions at the inner and outer margins of the graphite reflector have unsymmetrical shapes, similar to axial distributions in the core. There is a dissimilarity between the distribution curves at the inner margin and those at the outer margin of the reflector. b/ The radial distribution on the upper surface of the graphite reflector is measured and is described by the two-group neutron diffusion theory. The maximal value of the curve lies at the position of Rmax = 22.5 cm. c/ The distribution in the twenty water irradiation holes around the rotary specimen rack is obtained. (author). 3 refs., 5 figs., 1 tab

  1. Thermal behavior of the nuclear graphite waste generated from the decommissioning of the nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Hee Chul; Eun, Hee Chul; Lee, Dong Gyu; Cho, Yong Jun; Kang, Young Ae; Lee, Keun Woo; Oh, Won Jin [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    This study investigated the thermal behavior of the nuclear graphite waste generated from the decommissioning of the Korean nuclear research reactor. The first part study investigated the decomposition rate of the nuclear graphite waste up to 1000 .deg. C under various oxygen partial pressures using a thermo-gravimetric analyzer (TGA). Tested graphite waste sample not easily destroyed in the oxygen-deficient condition. However, the gas-solid oxidation reaction was found to be very effective in the presence of oxygen. No significant amount of the products of incomplete combustion was formed even in the limited oxygen concentration of 4 % O{sub 2}. The influence of temperature and oxygen partial pressure was evaluated by the theoretical model analysis of the thermo-gravimetric data. The activation energy and the reaction order of graphite oxidation were evaluated as 128 kJ/mole and 1.1, respectively. The second part of this study investigated the behavior of radioactive elements under graphite oxidation atmosphere using thermodynamic equilibrium model. {sup 22}Na, {sup 134}Cs and {sup 137}Cs were found be the semi-volatile elements. Since volatile uranium species can be formulated at high temperatures above 1050 .deg. C, the temperature of incinerator furnace should be minimized. Other corrosion/activation products, fission products and uranium were found to be the non-volatile species.

  2. Thermal behavior of the nuclear graphite waste generated from the decommissioning of the nuclear research reactor

    International Nuclear Information System (INIS)

    This study investigated the thermal behavior of the nuclear graphite waste generated from the decommissioning of the Korean nuclear research reactor. The first part study investigated the decomposition rate of the nuclear graphite waste up to 1000 .deg. C under various oxygen partial pressures using a thermo-gravimetric analyzer (TGA). Tested graphite waste sample not easily destroyed in the oxygen-deficient condition. However, the gas-solid oxidation reaction was found to be very effective in the presence of oxygen. No significant amount of the products of incomplete combustion was formed even in the limited oxygen concentration of 4 % O2. The influence of temperature and oxygen partial pressure was evaluated by the theoretical model analysis of the thermo-gravimetric data. The activation energy and the reaction order of graphite oxidation were evaluated as 128 kJ/mole and 1.1, respectively. The second part of this study investigated the behavior of radioactive elements under graphite oxidation atmosphere using thermodynamic equilibrium model. 22Na, 134Cs and 137Cs were found be the semi-volatile elements. Since volatile uranium species can be formulated at high temperatures above 1050 .deg. C, the temperature of incinerator furnace should be minimized. Other corrosion/activation products, fission products and uranium were found to be the non-volatile species

  3. Fractionalization of graphitic reactor components

    International Nuclear Information System (INIS)

    Within the scope of this project, a stock taking took place that comprised graphic components in German research and prototype reactors. From a long set of experiments mechanical breaking was determined to represent the optimal solution for the problem of fractionalizing graphitic components with a reasonable tritium content. Be sides the very low dust emission and the total lack of secondary waste emission, the small and mobile tool are outstanding advantages of this technique. Graphitic dust which was emitted throughout the different experiments was characterized and analyzed with respect to its explosion properties. Choosing the right conditions, a graphite dust explosion can safety be expelled during the trials on activated graphite. Successful fractionalization experiments performed on activated graphite clearly show that a planar mechanical breaking of graphite as well as a yawing of graphitic components is possible. The performance of the filter installation is evidenced by the fact that all tools and components - with exception of the filters themselves - applied during the experiments could be decontaminated and measured for free release. (orig.)

  4. Fractionalization of graphitic reactor components

    International Nuclear Information System (INIS)

    In Germany's research- and prototype reactors, some 590 t of graphite and carbonstone are in use, safe enclosure or due to decommissioning. Different concepts for their decommissioning are currently being discussed, all of which have to include any kind of fractionalization due to the fact that container dimensions do not refer to sizes of the above mentioned ceramic components. By means of parting techniques, activated/contaminated parts ought to be delaminated or parted in a way that a maximum ratio of container filling and the referring volume is reached. Thus, a maximum of final storage volume yield coinciding with a minimization of final disposal costs can be achieved. Fractionalization techniques will be developed or fitted with the aim of secondary waste minimization. Special focus ist put on the development of dust collection techniques in respect of emerging graphite dust out of the fractionalizating techniques. In 1995, a task group build up a concept regarding the dismantling of the Juelich AVR reactor to the green meadow [1]. One result of this study was the necessary amount containers for the final storage of all graphitic parts of the mentioned high-temperature reactor. The referred number was estimated to be 2511 MOSAIK II-type containers or 364 type VII-containers. To fill these containers, most of the graphitic components have to be parted. (orig.)

  5. Impact strength of reactor graphite

    International Nuclear Information System (INIS)

    Dose dependences of impact strength of reactor graphite irradiated with neutrons at 340-360, 420-460 and 620-800 K by the fluence of up to 1.3x1022 cm-2 are plotted. It is pointed out that for the temperature range and fluence studied impact strength, as well as crack resistance, of reactor graphite exceed the same parameters of graphite before irradiation

  6. BROOKHAVEN

    International Nuclear Information System (INIS)

    Neutrino physics has been an integral part of the Brookhaven research programme for much of the Laboratory's 46-year history. Milestones have been the determination of the helicity of neutrinos (1958), the establishment of the existence of two kinds of neutrinos (1962) for which Leon Lederman, Mel Schwartz and Jack Steinberger were awarded the 1988 Nobel Physics Prize, the discovery of charmed baryons in the 7' Bubble Chamber in 1975, and the ongoing measurements by Ray Davis and collaborators of solar neutrinos, first reported in 1968. There have also been significant contributions to the understanding of neutral currents in exclusive hadron and electron channels. In addition some of the earliest, and to date best, accelerator limits on electron muon neutrino oscillations are from Brookhaven experiments. The Laboratory is also the 'B' in the IMB underground experiment, built to search for proton decay and which caught neutrinos from the SN 1987a supernova. At present Brookhaven is heavily involved in the Gallex project in the Gran Sasso and recently a new collaboration has received scientific approval for a long baseline experiment to search for muon neutrino oscillations via muon neutrino disappearance

  7. Graphite reactor physics

    International Nuclear Information System (INIS)

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm2, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors)

  8. Biological efficiency of the Brookhaven Medical Research Reactor mixed neutron beam estimated from gene mutations in Tradescantia stamen hair cells assay

    International Nuclear Information System (INIS)

    The relative biological effectiveness (RBE) of low energy neutrons for the induction of various abnormalities in Tradescantia stamen hair mutation (Trad-SH) assay was studied using two clones (T-4430 and T-02), heterozygous for flower color. Dose response relationship for gene mutations induced in somatic cells of Trad-SH were investigated after irradiation with a mixed neutron beam of the Brookhaven Medical Research Reactor (BMRR), currently used in a clinical trial of boron neutron capture therapy (BNCT) for glioblastoma. To establish the RBE of the BMRR beam in the induction of various biological end-points in Tradescantia, irradiation with various doses of γ-rays was also performed. After irradiation all plants were cultivated several days at Brookhaven National Laboratory (BNL), then transported to Poland for screening the biological end-points. Due to the post-exposure treatment, all plants showed high levels of lethal events and alteration of the cell cycle. Plants of clone 4430 were more reactive to post-treatment conditions, resulting in decreased blooming efficiency that affected the statistics. Slope coefficients estimated from the dose response curves for gene mutation frequencies allowed the evaluation of ranges for the maximal RBE values of the applied beam vs. γ rays as 6.0 and 5.4 for the cells of T-02 and T-4430, respectively. Estimated fraction of doses from neutrons and corresponding biological effects for the clones T-02 and T-4430 allowed to evaluate the RBE values for neutrons part in the beam as 32.3 and 45.4, respectively. (author)

  9. Rebuilding the Brookhaven high flux beam reactor: A feasibility study

    International Nuclear Information System (INIS)

    After nearly thirty years of operation, Brookhaven's High Flux Beam Reactor (HFBR) is still one of the world's premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR's value as a national scientific resource, members of the Laboratory's scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor's research capabilities

  10. Incineration of graphite used in nuclear reactor

    International Nuclear Information System (INIS)

    Tokai Power Station was closed on March 31, 1998 and it is scheduled to be undertaken decommissioning. 1,600 tons of graphite is used in the nuclear reactor as moderator which was irradiated by neutron during the operation. Incineration of the graphite is the largest problem. From this background, graphite oxidation has been investigated, where an ideal method has been invented to practice perfect combustion of radioactive graphite. This new method has been patented on the fundamental simple process which comprises the use of oxygen and no needs of the graphite pulverization. (author)

  11. Incineration of graphite used in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuroyama, Shigefumi [Nuclear Services Co., Tokyo (Japan)

    2001-10-01

    Tokai Power Station was closed on March 31, 1998 and it is scheduled to be undertaken decommissioning. 1,600 tons of graphite is used in the nuclear reactor as moderator which was irradiated by neutron during the operation. Incineration of the graphite is the largest problem. From this background, graphite oxidation has been investigated, where an ideal method has been invented to practice perfect combustion of radioactive graphite. This new method has been patented on the fundamental simple process which comprises the use of oxygen and no needs of the graphite pulverization. (author)

  12. Techniques of graphite irradiation in nuclear reactors

    International Nuclear Information System (INIS)

    The chapter highlights design criteria of in-pile devices for irradiation of graphite samples, as well as techniques for determination and comparison of irradiation histories. The design and technological capabilities of capsules for irradiating samples in the MR and WWR-M research reactors at neutron fluxes (E >= 0.5 MeV) up to 2x1014 neutron/cm2xs and temperatures up to 1000 deg C are described. A short survey is also given of radiation damage mechanisms in graphite and of the role of the major factors - temperature and neutron fluence. In this connection modern methods of measuring in-core temperature and neutron fluences are considered. The correlations between different measuring units of fluence and equivalent fission fluence are provided

  13. New Brookhaven chief seeks cross-cutting research

    CERN Multimedia

    Jones, D

    2003-01-01

    Brookhaven National Laboratory will pursue opportunities for promoting commercial development of energy systems and other technologies while focusing on the lab's primary mission of basic science research, according to the incoming BNL director, Praveen Chaudhari (1 page).

  14. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  15. Radiochemical characterization of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Nuclear reactors which have in-built graphite may receive a high neutron dose for a long period. Depending on the chemical composition of the graphite, numerous activation products may result. In addition, the amount of fission product contamination will depend on the location of the graphite. The migration of fission products may be supported by the high temperatures which occur in high-temperature reactors. At the Juelich 15 MWe high-temperature gas-cooled experimental AVR (Arbeitsgemeinschaft Versuchsreaktor) reactor, two different types of nuclear graphite had been in use. High-purity graphite was used as a basic material for core structures of the AVR. Insulation layers of carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For various reasons it is important to know the degree of contamination of graphite and carbon bricks from activation and fission products. The optimum method for nuclear graphite analysis in decommissioning is by incineration. Volatile activities (14C, 3H, 36Cl, ...) have to be captured for analysis. In cases where dust-like samples are handled, the incineration furnace has to be small enough to be operated in a glove-box. The resulting ashes can be used for determining all non-volatile nuclides by different radiochemical methods. In early 1999, some graphite and carbon brick samples from the AVR reactor were obtained by drilling. The samples were then analysed in the laboratories at the Juelich research centre. For incineration a vertical quartz tube was used which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl were captured in washing bottles. After further preparation, they were analysed by liquid scintillation counting (LSC). After dissolving the ashes, the elements were separated by ion exchange, extraction methods and HPLC. The radionuclides were then determined by alpha-spectrometry, LSC, low

  16. Activity evaluation of 60Co, 137Cs, 152Eu, 154Eu, 14C and total β in graphite from the thermal column of Romanian TRIGA research reactor

    International Nuclear Information System (INIS)

    Several methods were investigated for identification and activity evaluation of 60Co, 137Cs, 152Eu, 154Eu, 14C and total β in graphite samples collected from the thermal column of Romanian research reactor. An ORTEC DIGIDART multichannel gamma spectrometer was used for measurement of 60Co, 137Cs, 152Eu and 154Eu. The main contribution to the radioactivity of graphite comes from 60Co, 152Eu and 154Eu. Combustion of graphite samples using a PerkinElmer sample oxidizer was utilized for total oxidizing of 14C to gaseous carbon dioxide. A TRI-CARB Packard liquid scintillation analyzer was used for 14C counting. The degree of 14C recovery in the analyzed standard samples was higher than 98%, and the 14C retention degree in the circuits of combustion system was less than 0.08%. To evaluation of total β activity in the irradiated graphite, the samples were measured using a proportional counter. The mean total β activity in the irradiated graphite samples was about 3000 Bq/g. (authors)

  17. Separation of Graphitic Line in Debyegram of the Reactor Graphite

    International Nuclear Information System (INIS)

    The author sets forth a method for the determination of crystallite dimensions Lc in reactor graphite. The method consists of the mathematical correction of the diffraction profile (002) obtained by means of a Philips diffractometer with scintillation counter. During the work CuKβ radiation is used. The convenience of using CuKβ radiation rather than that of a CuKα doublet is discussed. All instrumental errors except that of the sample adsorption are eliminated using corresponding slits. Sample adsorption is corrected mathematically. Highly symmetrical lines of graphitic and non-graphitic phase could be analytically separated from such corrected line profiles. Experimental error in the determination of the half-height and half-width of these lines does not exceed 5%. (author)

  18. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

  19. Acceptance test of graphite components in nuclear reactor

    International Nuclear Information System (INIS)

    The HTTR is the first high temperature gas-cooled reactor in Japan. It is a test reactor with thermal power of 30 MW and coolant outlet temperature of 950degC at maximum. To achieve high temperature coolant core internals were made of graphite and carbon materials due to their excellent thermal resistivity. After fabrication of graphite and carbon components at works they were installed in the HTTR, and now it is in the power up testing stage. Concerning the inspection standard of the graphite and carbon components, nondomestic standard exists as main components in the nuclear reactor. It is necessary, therefore, to prescribe the inspection standards for the HTTR graphite components. Many research and developments in relation to the inspection standard, e.g. in the research field of nondestructive examination of the graphite material, had been performed, and then the JAERI established the inspection standard. The acceptance test of the graphite and carbon components was carried out based on the inspection standard. This paper prescribes the outline of the established inspection standard. (author)

  20. Development of the epithermal neutron beam and its clinical application for boron neutron capture therapy at the Brookhaven medical research reactor

    International Nuclear Information System (INIS)

    The failures of the Boron Neutron Capture Therapy (BNCT) trials conducted between 1951 and 1961 were attributed to inadequate penetration of the thermal neutron beams and poor localization of boron compound in the tumour. The epithermal neutron beam at the BMRR was designed and installed to improve the penetration of the neutron beam. The use of this epithermal neutron beam for the clinical trial initiated in 1994 at Brookhaven National Laboratory (BNL) was preceded by the neutron beam optimization and characterization, the validation of the treatment planning software and the establishment of a procedure for treatment plan evaluation and dose reporting and recording. To date, a total of 54 patients have been treated. Our experience in the development of the epithermal neutron beam for clinical BNCT at the BMRR may be useful to other investigators desirous of developing similar programs for cancer therapy. (author)

  1. Radiochemical characterisation of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Graphite built-in nuclear reactors may receive a high neutron dose for a long period. Depending on its chemical composition a lot of activation products are produced. In addition, there is more or less fission product contamination depending on the location. The migration of fission products may be supported by high temperatures which occur in high temperature reactors. At the Juelich 15 MWe High Temperature Gas-cooled experimental Reactor AVR (Arbeitsgemeinschaft Versuchsreaktor) two different types of nuclear graphite had been in use. High-purity graphite was used as basic material for core structures of the AVR. Insulation layers from carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For many reasons it is important to know the amount of contamination of graphite and carbon bricks with activation products and fission products. The head end of nuclear graphite analytics must be the incineration. Volatile activities (14C, 3H, 36Cl ...) must be caught for determination. In case of handling dustlike samples the incineration furnace must be small enough to be operated in a glove box. The resulting ashes can be used for determining all non volatile nuclides with different radiochemical methods. In early 1999 some graphite and carbon brick samples from AVR-reactor had been taken by drilling. The samples had been analysed in our laboratories at Juelich research centre. For incineration we used a vertical quartz-tube which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl are caught in washing bottles. After further preparation, they are determined by LSC. After dissolving the ashes, the elements are separated by ion exchange, extraction methods and HPLC. The radionuclides are then determined by a-spectrometry, LSC, low level g-spectrometry and x-ray spectrometry. (author)

  2. US graphite reactor D ampersand D experience

    International Nuclear Information System (INIS)

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)

  3. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Brookhaven highlights. Report on research, October 1, 1992--September 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Belford, M.; Cohen, A.; Greenberg, D.; Seubert, L. [eds.

    1993-12-31

    This report highlights the research activities of Brookhaven National Laboratory during the period dating from October 1, 1992 through September 30, 1993. There are contributions to the report from different programs and departments within the laboratory. These include technology transfer, RHIC, Alternating Gradient Synchrotron, physics, biology, national synchrotron light source, applied science, medical science, advanced technology, chemistry, reactor physics, safety and environmental protection, instrumentation, and computing and communications.

  5. Brookhaven highlights

    International Nuclear Information System (INIS)

    This report highlights the research activities of Brookhaven National Laboratory during the period dating from October 1, 1992 through September 30, 1993. There are contributions to the report from different programs and departments within the laboratory. These include technology transfer, RHIC, Alternating Gradient Synchrotron, physics, biology, national synchrotron light source, applied science, medical science, advanced technology, chemistry, reactor physics, safety and environmental protection, instrumentation, and computing and communications

  6. Research reactors

    International Nuclear Information System (INIS)

    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

  7. Research reactors

    International Nuclear Information System (INIS)

    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)

  8. Examination of graphite pile radiation conditions of industrial reactors

    International Nuclear Information System (INIS)

    Radiation states of graphite piles of three industrial uranium-graphite reactors are investigated. The level, composition, distribution of the pile radioactive contamination, parameters of neutron and gamma radiations are determined. The forecast of variation of radionuclide activity in graphite in dependence on the cooling time is developed

  9. Brookhaven highlights - Brookhaven National Laboratory 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report highlights research conducted at Brookhaven National Laboratory in the following areas: alternating gradient synchrotron; physics; biology; national synchrotron light source; department of applied science; medical; chemistry; department of advanced technology; reactor; safety and environmental protection; instrumentation; and computing and communications.

  10. Temperature distribution in graphite during annealing in air cooled reactors

    International Nuclear Information System (INIS)

    A model for the evaluation temperature distributions in graphite during annealing operation in graphite. Moderated an-cooled reactors, is presented. One single channel and one dimension for air and graphite were considered. A numerical method based on finite control volumes was used for partioning the mathematical equations. The problem solution involves the use of unsteady equations of mass, momentum and energy conservation for air, and energy conservation for graphite. The source term was considered as stored energy release during annealing for describing energy conservation in the graphite. The coupling of energy conservation equations in air and graphite is performed by the heat transfer term betwen air and graphite. The results agree with experimental data. A sensitivity analysis shown that the termal conductivity of graphite and the maximum inlet channel temperature have great effect on the maximum temperature reached in graphite during the annealing. (author)

  11. Actinides in irradiated graphite of RBMK-1500 reactor

    International Nuclear Information System (INIS)

    Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK

  12. Safety regulations for the nuclear uranium-graphite production reactors decommissioning in the Russian Federation

    International Nuclear Information System (INIS)

    There are different types of nuclear installations in the Russian Federation. Among them there are: nuclear power plants (NPPs), research reactors and critical facilities, fuel cycle facilities including uranium mining, conversion, enrichment, nuclear fuel fabrication, spent nuclear fuel (SNF) reprocessing, SNF and radioactive waste (RAW) storage facilities, radiochemistry, production reactors, etc. The first experience in the nuclear fuel cycle facilities decommissioning in Russia was gained from the uranium graphite production reactors shutdown at 'Mayak' site (Ozersk, Chelyabinsk region). The first uranium-graphite reactor was shutdown in 1987 at 'Mayak' site (it was the research uranium-graphite reactor AI). Sometime later in that year the next production reactor (A) was shutdown at the same site. This decommissioning project was used as the prototype for the decommissioning projects of the next two production reactors (AV-2 and AV-3) at the 'Mayak' site (both were shutdown in 1990). In the same year two uranium-graphite production reactors (I1 and EI-2) were shutdown at the 'SChC' site (Seversk, Tomsk region). In 1992 three uranium-graphite production reactors were shutdown - one (ADE-3) at the 'SChC' site and two (AD and ADE-1) at 'MChC' site (Zeleznogorsk, Krasnoyarsk region). (author)

  13. Job and Task Analysis project at Brookhaven National Laboratory's high flux beam reactor

    International Nuclear Information System (INIS)

    The presenter discussed the Job and Task Analysis (JTA) project conducted at Brookhaven National Laboratory's High Flux Beam Reactor (HFBR). The project's goal was to provide JTA guidelines for use by DOE contractors, then, using the guidelines conduct a JTA for the reactor operator and supervisor positions at the HFBR. Details of the job analysis and job description preparation as well as details of the task selection and task analysis were given. Post JTA improvements to the HFBR training programs were covered. The presentation concluded with a listing of the costs and impacts of the project

  14. Brookhaven highlights

    International Nuclear Information System (INIS)

    This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future

  15. Brookhaven highlights

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. (eds.)

    1992-01-01

    This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future.

  16. Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

    International Nuclear Information System (INIS)

    Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on

  17. Determination of Carbon-14 (14C) and Tritium (3H) in the graphite thermal column of a IRT-2000 research reactor in Sofia

    International Nuclear Information System (INIS)

    Radiocarbon has average, and tritium has low level of radiotoxicity, but both are easily absorbable radionuclide from the body as due to their chemical nature they are included in almost all the chemical compounds involved in metabolic chains and in structure of proteins and biological macromolecules. Due to above mentioned reasons, the analysis and determination of these two major radionuclides is very significant for radiation protection for both the public and the environment as well as for occupational exposure of personnel. This publication describes the detailed procedure for determining the concentration of 3H and 14C in samples selected at the stage 'partial dismantling' of the research nuclear reactor IRT-2000 reconstruction. The procedure consists graphite sample thermal burning in the tubing muffle furnace at a maximum temperature of 850'0C. After combustion the gas mixture passes through a CuO catalyst, heated to 450-5000C and through a series of traps for 3H and 14C capturing. The captured CO2+ 14CO2 as Na2CO3 in NaOH precipitates with BaCl2 as BaCO3 and it is followed by a liquid scintillation spectrometry measurements

  18. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  19. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  20. Method to manufacture graphite forms for atomic nuclear reactors

    International Nuclear Information System (INIS)

    Graphitic forms can be manufactured from petroleum-based or coal-tar based coke which still contains 8-20% volatile components. This softens at 400-5500C and the crystal structure is disordered. One obtains an isotropic coke as far as the thermal expansion coefficient is concernend. The addition of softeners (coal tar oils, pitch, phenantrene etc.) also leads to isotropy. This pretreated coke is formed (addition of binder), burnt and graphitized. These graphite forms are used in nuclear reactors as they are very stable and have almost the same heat expansion coefficients in all three directions (4-5 x 10-6/0C). (IHOE)

  1. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    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

  2. Investigation of graphite pile radiation features of uranium-graphite reactor AI

    International Nuclear Information System (INIS)

    Paper presents the results of the examination of the Mayak PA uranium-graphite AI reactor stack. On the basis of the mentioned results one drew up the report on the stack nuclear safety and the radiation certificate. The radiation examination enabled to determine the level, the composition and the distribution of the stack contamination and the distribution of the neutron and γ-radiation, to predict variation of radionuclide activity within the graphite depending on the cooling time. The mentioned data are necessary to ensure safety analysis and to make decisions on further stages of the reactor decommissioning

  3. Radiation annealing of reactor graphite after neutron irradiation

    International Nuclear Information System (INIS)

    The reactor graphite specimens initially irradiated at 310-450 K by the fluences 5x1021, 9x1021, 1,3x1022 n/sm2 (E>0,18 MeV) were repeatedly irradiated at 620-1200 K by the fluences up to 2,5x1021 n/sm2. Simultaneously the control specimens of reactor graphite and nongraphitized half finished product were irradiated. It is shown that irradiation annealing of preirradiated specimens leads to the more considerable shrinkage that of the control specimens

  4. Nuclear research reactors

    International Nuclear Information System (INIS)

    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.)

  5. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  6. Influence of production technology on reactor graphites irradiation resistance

    International Nuclear Information System (INIS)

    The results of comparative service life tests of samples of the industrially produced graphites GR-280, ARV, MPG-6, as well as process variants of experimental graphites with pilot grades GR-1 and SPP-MPG-R based on composite filler are discussed. The tests are conducted in the BOR-60 reactor lateral shield at temperature of 350-450 deg C. The data characterizing changes in size, thermal expansion factor and dynamic Young modulus for the samples studied are presented. It is shown that radiation resistance of graphite increases in the series SPP-MPG-R → GR-280 → ARV → GR-1, MPG-6 which is the result of anisotropy properties decrease, strength improvement and graphite structure homogenization in initial state. The conclusion is drawn that the critical fluence corresponding to the Young modulus decreasing up to the initial value is the most adequate characteristic of the material behaviour under irradiation

  7. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  8. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Progressive Application Decommissioning Models for U.S. Power and Research Reactors

    International Nuclear Information System (INIS)

    This paper presents progressive engineering techniques and experiences in decommissioning projects performed by Bums and Roe Enterprises within the last fifteen years. Specifically, engineering decommissioning technical methods and lessons learned are discussed related to the Trojan Large Component Removal Project, San Onofre Nuclear Generating Station (SONGS) Decommissioning Project and the Brookhaven Graphite Research Reactor (BGRR) Decommissioning Project Study. The 25 years since the 1979 TMI accident and the events following 9/11 have driven the nuclear industry away from excessive, closed/elitist conservative methods towards more pragmatic results-oriented and open processes. This includes the essential recognition that codes, standards and regulatory procedures must be efficient, effective and fit for purpose. Financial and open-interactive stakeholder pressures also force adherence to aggressive risk reduction posture in the area of a safety, security and operations. The engineering methods and techniques applied to each project presented unique technical solutions. The decommissioning design for each project had to adopt existing design rules applicable to construction of new nuclear power plants and systems. It was found that the existing ASME, NRC, and DOE codes and regulations for deconstruction were, at best, limited or extremely conservative in their applicability to decommissioning. This paper also suggests some practical modification to design code rules in application for decommissioning and deconstruction. The representative decommissioning projects, Trojan, SONGS and Brookhaven, are discussed separately and the uniqueness of each project, in terms of engineering processes and individual deconstruction steps, is discussed. Trojan Decommissioning. The project included removal of entire NSSS system. The engineering complexity was mainly related to the 1200 MW Reactor. The approach, process of removal, engineering method related to protect the worker

  10. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  11. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  12. Radiation characteristics of graphite in uranium-graphite reactors after decommissioning

    International Nuclear Information System (INIS)

    The results of calculating the specific activity of long-living radioisotopes in reactor graphite and uranium fuel are discussed. For applied estimations the data are normalized for operational parameters comparatively general for all reactors with graphite moderators and uranium fuel which are the following: irradiation period is 30 years, thermal neutron flux density is 3 x 1013 s-1 x cm-2, spectrum hardness is 0.05, temperature is 900 K. It is assumed that radioisotopes produced in the course of operation do not migrate through the masonry.It is shown that radioactivity for long-living beta decay radioisotopes is proportional to neutron fluence in the first approximation. The conclusion is made that the local exposure γ radiation dose rate in the masonry will amount to approximately 102 - 103 R/h even after cooling for 100 years

  13. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    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

  14. Advanced pebble bed high temperature reactor with central graphite column for future applications

    International Nuclear Information System (INIS)

    Design evaluations of the advanced pebble bed high temperature reactor, AHTR, with central graphite column are given. This reactor, as a nuclear heat source, is suitable for coal refinement as well as for electricity generation with closed gas turbine primary helium circuit. With this design of the central graphite column, it is possible to limit the core temperatures under the required value of about 1600deg C in case of accident conditions, even with higher thermal power and higher core inlet and outlet temperatures. The designs of core internals are described. The after heat removal system is integrated in the prestressed concrete reactor pressure vessel, which is based on the principals of natural convection. Research work is being carried out, whereby the sphencal fuel elements are coated with a layer of silicon carbide, to improve the corrosion resistance as well as the effectiveness of the fission products barrier. (orig.)

  15. Survey of research reactors

    International Nuclear Information System (INIS)

    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)

  16. Safeguarding research reactors

    International Nuclear Information System (INIS)

    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

  17. Research nuclear reactors

    International Nuclear Information System (INIS)

    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.)

  18. Utilization of water-graphite reactor laying material as a secondary raw one

    International Nuclear Information System (INIS)

    Problem on utilization of spent graphite laying material of water-graphite reactors is considered. The spent graphite is a valuable raw material, application where of makes it possible to obtain composition material with high physical-mechanical characteristics. Possible application areas of the above composition materials with modified irradiation by a filler are indicated. 5 refs.; 3 tabs

  19. Design guide for category VI reactors: air-cooled graphite reactors

    International Nuclear Information System (INIS)

    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 air-cooled graphite 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

  20. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    G1 (46 MWth), G2 (250 MWth) and G3 (250 MWth) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide (14C, 36Cl, 63Ni, 60Co,3H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  1. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors)

  2. Multipurpose research reactors

    International Nuclear Information System (INIS)

    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

  3. Graphite Technology Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; T. Burchell; M.Carroll

    2010-10-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled High Temperature Gas Reactor (HTGR) with a large graphite core. Graphite physically contains the fuel and comprises the majority of the core volume. Graphite has been used effectively as a structural and moderator material in both research and commercial high-temperature gas-cooled reactors. This development has resulted in graphite being established as a viable structural material for HTGRs. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermomechanical design of the structural graphite in NGNP is based. This Technology Development Plan outlines the research and development (R&D) activities and associated rationale necessary to qualify nuclear grade graphite for use within the NGNP reactor.

  4. Research reactors in Argentina

    International Nuclear Information System (INIS)

    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

  5. Ageing of research reactors

    International Nuclear Information System (INIS)

    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)

  6. Reactor Materials Research

    International Nuclear Information System (INIS)

    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)

  7. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

    International Nuclear Information System (INIS)

    The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of 13C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions. The generation of carbon isotopes 13C and 14C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ13C/12C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 x 1014 n/cm2 s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of 14C is compared with the one measured by the β spectrometry technique. Results indicate that production of 14C from 14N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of 14C activity. Measured 14C specific activity values correspond to 15 ± 4 ppm impurity of 14N in graphite samples from the RBMK-1500 reactor core.

  9. Technical Safety Requirement Violation at the High Flux Beam Reactor Decommissioning Project, Brookhaven, United States of America

    International Nuclear Information System (INIS)

    At Brookhaven National Laboratory (BNL) on 6 July 2009, a technical safety requirement (TSR) violation was declared at the high flux beam reactor (HFBR) project, which was a limited scope decontamination and decommissioning project associated with the permanently shutdown reactor. The violation extended from performing decommissioning activities within the facility under the incorrect mode. The draining of the spent fuel pool was performed in the warm standby mode when it should have been in the operation mode. The TSR was developed contrary to the United States Department of Energy (DOE) TSR guidance, which recommends that facility operations should only be carried out in the operation mode. The facility TSR allowed operations to be carried out in both modes. The HFBR operation mode focused on the removal of a small number of highly irradiated components with associated limited conditions of operation (LCO), while the warm standby mode focused on all other tasks in the facility and did not require entry into the LCO

  10. Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

    International Nuclear Information System (INIS)

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR) as moderator, thermal absorber and also as structural components of the core. This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example, studies about the

  11. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    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

  12. Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H.

    1991-12-31

    This report highlights Brookhaven National Laboratory`s activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

  13. Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H.

    1991-01-01

    This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)

  14. Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

    International Nuclear Information System (INIS)

    This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed

  15. The replacement research reactor

    International Nuclear Information System (INIS)

    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

  16. Neutronic flux stability of production uranium graphite reactor conversion core relative to high-frequency oscillations

    International Nuclear Information System (INIS)

    Preliminary methodical simplified investigation into stability of the neutron field in the conversion load of industrial uranium-graphite reactors with regard to basic characteristics of the load in transient processes was carried out. Analysis was based on the calculated research into the behaviour of simplified single-point and one-dimensional models of the reactor core in transient regimes during the interconnected description of dynamics of neutron-physical and thermal properties of the load. Fundamental assumptions on the reactor characteristics used in the calculated model. In the context of accepted approximations the obtained results preclude the possibility for the occurrence of spontaneous high frequency oscillations resulting from the positive reactivity effect on the fuel temperature in the conversion load

  17. Brookhaven highlights, October 1979-September 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety). (GHT)

  18. Brookhaven highlights, October 1979-September 1980

    International Nuclear Information System (INIS)

    Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety)

  19. Void-free filling of reactor space cavities during uranium-graphite reactors decommissioning

    International Nuclear Information System (INIS)

    Uranium-graphite reactors (UGR) decommissioning is to be assured through reliable isolation of radioactive waste on the reactor site, providing for radiation protection of personnel, the public and the environment throughout the entire time that the waste will remain hazardous. The paper shows stages of additional safety barriers construction during UGR decommissioning using safe in-situ entombment. The technology of selecting material for man-made geo-barriers and the void-free filling technology and its mock-up trials are described

  20. Safety of research reactors

    International Nuclear Information System (INIS)

    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

  1. Research reactor support

    International Nuclear Information System (INIS)

    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. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235U or 239Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  3. Evaluation of research reactors

    International Nuclear Information System (INIS)

    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. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  5. Thai research reactor

    International Nuclear Information System (INIS)

    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)

  6. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    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

  7. TRIGA research reactors

    International Nuclear Information System (INIS)

    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.)

  8. Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

    International Nuclear Information System (INIS)

    The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)

  9. Dossier: research reactors

    International Nuclear Information System (INIS)

    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.)

  10. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    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

  11. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  12. First Algerian research reactor

    International Nuclear Information System (INIS)

    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

  13. The replacement research reactor

    International Nuclear Information System (INIS)

    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

  14. Concerning the choice of graphite for stacking of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Radiation behavior is considered of several forms of graphite in order to assess the service life of the structural elements of the reactor core. Above all, the dimensional stability of the graphite blocks should be ensured so as to eliminate the possibility of their splitting under contraction and breaking under secondary swelling. Results of the radiation-induced changes are presented in the dimensions of specimens of some grades of graphite: the American anisotropic CSF, isotropic RP4, and graphite based on gilsonite coke, and Soviet GMZ graphite and its variant, in which impregnation with pitch and graphitization at 2800/degree/C have been employed. By using high-density, high-strength isotropic carbon material it is possible to solve the problem of fatigue service life, radiation stability, and corrosion. 11 refs

  15. Definition gas isolation characteristics and hydrogen mass-transport in graphite based reactor materials

    International Nuclear Information System (INIS)

    The purpose of this work is study of hydrogen mass-transport and graphite materials sorption properties under different reactor conditions in hydrogen medium for prognosis its capacity for work by reactor accident . As samples for study were given graphite materials; reactor graphite with 7% titanium additive (RG-T) and pyrolytic graphite with low density (PGL). Irradiation in the hydrogen medium has been conducted in IVG.IM reactor. Experimental conditions for RG-T sample: irradiated temperature 693 K; hydrogen pressure 105 Pa, time 2 hours, total fluence 1.5·1015 neutron/cm2. Samples of pyre graphite were cut out from thermal assembly shell, which worked for t=4000 c in the reactor IVG.1M in the hydrogen flow under 60·105 Pa pressure and temperature 1923 K. Total fluence is equals 8·1017 neutron/cm2. Dependence of gas isolation from RG-T and PGL graphite under isochronous annealing and programming linear heating by 20-2000 K temperature interval with quantity isolated gas mass-spectrometric registration is received. There are tree peaks of gas insolation on thermo sorption curves under regime of linear heating. First peak is responsible for hydrogen sorbed on surface, second - in pores, and third - hydrogen gas isolation from material's volume: lattice defects, pore's inside surface, lattice diffusion from graphite grain. Diffusion coefficient by pores in the graphite RG-T and PGL are approximately equal D=2·10-6 cm2/s under 1023 K. Values of hydrogen diffusion coefficient from irradiated and non-irradiated volumes of the both graphite samples are compared

  16. Reactor Materials Research

    International Nuclear Information System (INIS)

    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

  17. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. An explication of design data of the graphite structural design code for core components of High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    The core graphite material used for the core graphite components such as fuel blocks, control rod blocks etc. of High Temperature Engineering Test Reactor (HTTR) is nuclear grade fine-grained isotropic graphite (IG-110). The material data on properties such as tensile strength, compressive strength, Young's modulus, thermal expansion etc., including the effect of irradiation on these properties have been acquired up to now. From these obtained data the design data were stipulated by the graphite structural design code for core graphite components of HTTR in order to design the core graphite structures. This report explicates these design data for the core graphite components of the HTTR. (author)

  19. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type

    International Nuclear Information System (INIS)

    In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)

  20. Gaseous fuel reactor research

    Science.gov (United States)

    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.

  1. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  2. Bulk-bronzied graphites for plasma-facing components in ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600 degree C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000 degree C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800 degree C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350 degree C. 38 refs., 5 figs

  3. PROTEUS research reactor

    International Nuclear Information System (INIS)

    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

  4. Calculation of reactivity of control rods in graphite moderated reactors

    International Nuclear Information System (INIS)

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies

  5. Design of Modern Reactors for Synthesis of Thermally Expanded Graphite

    OpenAIRE

    Strativnov, Eugene V.

    2015-01-01

    One of the most progressive trends in the development of modern science and technology is the creation of energy-efficient technologies for the synthesis of nanomaterials. Nanolayered graphite (thermally exfoliated graphite) is one of the key important nanomaterials of carbon origin. Due to its unique properties (chemical and thermal stability, ability to form without a binder, elasticity, etc.), it can be used as an effective absorber of organic substances and a material for seal manufacturi...

  6. TSX graphite for extended use in the N-Reactor

    International Nuclear Information System (INIS)

    This report reviews the limited amount of irradiation data available for grade TSX graphite with the purpose of obtaining reasonable estimates of material behavior. The results are enhanced by obtaining generalized behavior characteristics demonstrated by similar grades of graphite, such as CSF, AGOT, and PGA. Intent of this work is to furnish the necessary coefficients to describe the material behavior for inclusion in the constitutive equations for the anisotropic graphite grade TSX. Estimates of the free-dimensional changes of TSX graphite as a function of temperature and fluence have been made and shown to be in good agreement with the data. The effects of irradiation on other physical properties, such as elastic moduli, conductivity, and coefficient of thermal expansion, are also described. The irradiation creep characteristics of TSX graphite are also estimated on the basis of data for similar grades of graphite in the US and Europe. Crude approximations of stresses generated in the keyed structure were made to demonstrate the magnitude of the problem. The results clearly predict that the filler-block keys will fail and the tube-block keys will not. It is also indicated that the overall stack height growth will be increased by 25 to 38 mm (1 to 1.5 in.) because of creep

  7. Fast breeder reactor research

    International Nuclear Information System (INIS)

    reactors of the future, the body of research aimed at developing liquid metal cooled fast reactors, national plans for work in 1976 on developing fast reactors - these were some of the topics discussed in connection with the national programmes. The development of power reactors involves a wide range of problems in the fields of nuclear and reactor physics, the thermophysics, chemistry, physics and technology of the cooling system, structural materials and nuclear fuel, the fabrication of reliable fuel elements and operating equipment, reactor monitoring and control, spent fuel reprocessing, the economics of constructing fast power reactors, nuclear safety, etc. The IWGFR, as at previous meetings, therefore paid great attention to the matter of holding international specialists' meetings. The working group recommended that the IAEA should organize the following IWGFR meetings in 1976: (1) In-Service Inspection and Monitoring (Bensberg, FRG, March 1976). (2) Cavitation in Sodium and Studies of Analogy with Water as Compared to Sodium (Cadarache, France, April 1976). (3) High Temperature Structural Design Technology (United States, May 1976) (4) Aerosol Formation, Vapour Deposits and Sodium Vapour Trapping (France, September-December 1976). The Group welcomed the IAEA's proposal to hold specialists' meetings on 'Fast Reactor Instrumentation' and 'Fuel Reprocessing and Recycling Techniques' within the framework of the Agency's programme of working groups in 1976. After discussing questions of co-ordinating and organizing international conferences on fast reactors, the IWGFR agreed to send representatives to the joint meeting of the American Nuclear Society and the American Institute of Metallurgical Engineers on 'Liquid Metal Technology', to be held at Champion, Pennsylvania, U.S.A. from 3-6 May 1976, and recommended that the IAEA should organize an international symposium on the 'Design, Construction and Operating Experience of Demonstration Fast Power Reactors' at Bologna

  8. Degradation of graphite in gas cooled reactors due to radiolytic oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Moskovic, R., E-mail: robert.moskovic@magnoxnorthsites.com

    2014-04-01

    Magnox reactors employ pile grade A (PGA) graphite as a moderator. Reactor cores are constructed typically of twelve to thirteen layers of graphite bricks. Fuel channels (FC) are in the centre of all bricks and interstitial channels (IC) at the centre of the corners of every second set of four bricks. The reactor core is cooled by carbon dioxide, the temperature of graphite core increases from 250 °C at the bottom to 360 °C at the top of the core. The neutron dose increases progressively with the operating time of the reactor. The graphite core looses mass as a result of radiolytic oxidation. The process is dependent on both total energy deposition and temperature which correlates with core height. Fast neutron dose accumulates at the same rate as the total energy deposited and is readily available. The reduction of density of moderator graphite increases the porosity and in turn changes both the physical and mechanical properties of graphite. The mechanical properties and density of graphite are measured either on samples installed in the reactor prior to service or trepanned from graphite bricks. The data obtained on these samples are interrogated using probability modelling to establish trends with increasing service life. Results of the analyses are illustrated in the paper. PGA graphite is an aggregate of coarse needle coke filler particles within a matrix of fine coke flour particles mixed with pitch binder. The bricks are fabricated in the green condition by extrusion of dry calcinated coke impregnated with liquid pitch binder and then graphitized at 2800 °C. This produces a polygranular aggregate with orthotropic properties. The strength properties of graphite are measured using different types of tests. The most commonly used tests involve bending, uniaxial and diametral compression. The initiation and propagation of cracks was investigated to improve understanding of strength behaviour. Cracking was examined on macro-scale using optical microscopy and

  9. Research Reactors of Ukraine

    International Nuclear Information System (INIS)

    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. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    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

  11. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  12. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  13. Applications of Research Reactors

    International Nuclear Information System (INIS)

    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

  14. Department of Energy’s ARM Climate Research Facility External Data Center Operations Plan Located At Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Cialella, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Gregory, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Lazar, K. [Brookhaven National Lab. (BNL), Upton, NY (United States); Liang, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ma, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Tilp, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Wagener, R. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-05-01

    The External Data Center (XDC) Operations Plan describes the activities performed to manage the XDC, located at Brookhaven National Laboratory (BNL), for the Department of Energy’s Atmospheric Radiation Measurement (ARM) Climate Research Facility. It includes all ARM infrastructure activities performed by the Data Management and Software Engineering Group (DMSE) at BNL. This plan establishes a baseline of expectation within the ARM Operations Management for the group managing the XDC.

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

    International Nuclear Information System (INIS)

    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

  16. Type A Verification Report For The High Flux Beam Reactor Stack And Grounds, Brookhaven National Laboratory, Upton, New York DCN: 5098-SR-08-0

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA).

  17. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    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

  18. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    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)

  19. Graphite-moderated and heavy water-moderated spectral shift controlled reactors

    International Nuclear Information System (INIS)

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

  20. Diffusion of Barium, Strontium and Cerium in Various Grades of Reactor Graphite

    International Nuclear Information System (INIS)

    Experiments designed to study the diffusion of barium, strontium and cerium in various grades of graphite of interest to the Dragon reactor project will be described. In these experiments specimens of the graphite are immersed in the vapour of the fission product metal under controlled conditions of temperature and pressure for known periods of time. After appropriate grinding and polishing, the distribution of the solute metal in the specimen is determined by means of an electron microprobe. The electron microprobe recordings are then interpreted to obtain apparent values of the diffusion coefficients and their dependence upon temperature. The results indicate that migration occurs primarily via defects in the graphite structure, with only a negligible contribution from in-grain diffusion. The significance of the results is discussed, with particular reference to the differences in behaviour encountered between various grades of graphite. (author)

  1. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Albarhoum, M., E-mail: pscientific1@aec.org.s [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)

    2011-01-15

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm{sup 3}. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  2. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    International Nuclear Information System (INIS)

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm3. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  3. Sorption of caesium and strontium by graphite materials in gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    The experiments have revealed that coked phenol resin binder has got an extremely high sorption capacity for Cs and Sr. For this reason the sorption capacity of A3 matrix graphite for fuel elements exceeds the capacity of the highly graphitized material that does not contain this component. The strong chemical binding for Cs and Sr by chemisorption indicates a retention of these elements when the nucleus is heated up by accident. The release calculations carried out with definite sorption isotherms revealed a larger retention effect by sorption for Sr than for Cs. In this respect the matrix graphite in the ball-shaped fuel elements is of special importance for the retention. It is applied at German high temperature reactors and contains non-graphitized phenol resin binder. (orig./DG)

  4. Experimental and numerical researches on collision of graphite bricks

    International Nuclear Information System (INIS)

    Highlights: • Collisions between two graphite bricks have been studied in the experiment. • Linear track and optics acquisition system were used to obtain collision characters. • The contact time decreases while the coefficient of restitution rises up with speed. • Specimen's rotating slightly to adjust posture is the main factor of the contact time. • The collision characters are sensitive to the posture deviation of specimens. - Abstract: Graphite bricks have important applications in high temperature gas-cooled reactors (HTGRs). The core of HTGR is a pebble-bed surrounded by graphite bricks, so vibrations and collisions between graphite bricks caused by external excitation have important influence on structural stability of the core. The experiment is aimed at studying the collisions between two bricks, using linear track test and optics acquisition system to obtain the collision characteristics of graphite bricks. Experimental results show the contact time decreases while the coefficient of restitution rises up with collision speed increasing. The results data are well-distributed and basic rules are credible. In the experiment, the impact surfaces did not make contact at the same time, as the specimens would twirl gently to adjust posture, and perfect flat collision was difficult to achieve. We simulated the collision process by a two-dimensional numerical model which had a small collision angle between the two contact surfaces. The simulation results fit well with experimental data and reveal the main influence factor of the collision characteristics. We have also discussed factors of small collision angle in the experiment, the dispersion degree of the contact time data in different speed ranges and the relationship between the coefficient of restitution and the speed. Discussion on the posture adjustment in a flat collision is an important part in the paper, and the analysis of that is meaningful for the future dynamic simulation of the core

  5. Experimental and numerical researches on collision of graphite bricks

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Lie; Wang, Hongtao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Sun, Libin, E-mail: slb@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Wang, Haitao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Ma, Shaopeng [Department of Mechanics, School of Aerospace Engineering, Beijing Institute of Technology, Beijing 100084 (China)

    2014-08-15

    Highlights: • Collisions between two graphite bricks have been studied in the experiment. • Linear track and optics acquisition system were used to obtain collision characters. • The contact time decreases while the coefficient of restitution rises up with speed. • Specimen's rotating slightly to adjust posture is the main factor of the contact time. • The collision characters are sensitive to the posture deviation of specimens. - Abstract: Graphite bricks have important applications in high temperature gas-cooled reactors (HTGRs). The core of HTGR is a pebble-bed surrounded by graphite bricks, so vibrations and collisions between graphite bricks caused by external excitation have important influence on structural stability of the core. The experiment is aimed at studying the collisions between two bricks, using linear track test and optics acquisition system to obtain the collision characteristics of graphite bricks. Experimental results show the contact time decreases while the coefficient of restitution rises up with collision speed increasing. The results data are well-distributed and basic rules are credible. In the experiment, the impact surfaces did not make contact at the same time, as the specimens would twirl gently to adjust posture, and perfect flat collision was difficult to achieve. We simulated the collision process by a two-dimensional numerical model which had a small collision angle between the two contact surfaces. The simulation results fit well with experimental data and reveal the main influence factor of the collision characteristics. We have also discussed factors of small collision angle in the experiment, the dispersion degree of the contact time data in different speed ranges and the relationship between the coefficient of restitution and the speed. Discussion on the posture adjustment in a flat collision is an important part in the paper, and the analysis of that is meaningful for the future dynamic simulation of the core

  6. Treatment of Irradiated Graphite to meet Acceptance Criteria for Waste Disposal: A New IAEA Collaborative Research Program - 12443

    International Nuclear Information System (INIS)

    World-wide, more than 250,000 tonnes of irradiated graphite have arisen through commercial nuclear-power operations and from military production reactors. Whilst most nations responsible for the generation of this material have in mind repository disposal alongside other radwaste, the lack of progress in this regard has led in some cases to difficulties where, for example, the site of an existing graphite-moderated reactor is required for re-utilisation. In any case, graphite as a radwaste stream has unique chemical and physical properties which may lend itself to more radical and innovative treatment and disposal options, including the recovery of useful isotopes and also recycling within the nuclear industry. Such aspects are important in making the case for future graphite-moderated reactor options (for example, High-Temperature Reactors planned for simultaneous power production and high-grade heat sources for such applications as hydrogen production for road fuel). A number of initiatives have taken place since the mid 1990s aimed at exploring such alternative strategies and, more recently, improving technology offers new options at all stages of the dismantling and disposal process. A new IAEA Collaborative Research Program aims to build upon the work already done and the knowledge achieved, in order to identify the risks and uncertainties associated with alternative options for graphite disposal, along with cost comparisons, thus enabling individual Member States to have the best-available information at their disposal to configure their own programs. (authors)

  7. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    CERN Document Server

    Mirfayzi, S R

    2013-01-01

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

  8. Neutron energy spectrum in graphite blankets of fusion reactors

    International Nuclear Information System (INIS)

    Neutron flux measurements were performed in a graphite stack and compared with calculations made with a two dimensional transport computer code. In the present work it is observed that the calculated spectrum in the elastic and inelastic scattering ranges (the first collision range in both cases), is sensitive to details of the angular distribution of these neutrons. Regarding the discrepancies in the elastic scattering range it is concluded that the microscopic cross section library ENDF/B-IV overestimates the large angle scattering (back scattering) as can be seen from comparison of measured and calculated spectra. The two most important conclusions of the present work are: 1. Inelastic scattering interaction of D-T neutrons in graphite cannot be calculated without a proper account of energy-angle correlation. 2. An experimental setup supplying monoenergetic collimated D-T neutrons constitutes a sensitive although indirect means for measuring angular distributions in inelastic and elastic scattering

  9. Present status of graphite-moderated power reactor decommissioning in foreign countries

    International Nuclear Information System (INIS)

    From 1960's on, graphite-moderated power reactors, being either of CO2 gas cooled or light water cooled type, had opened the nuclear electricity generation worldwide. Such pioneering reactors as UK Magnoxes, French GCRs, Russian AMBs had been operated for more than 20 years up to 40 years. Some of these pioneering power reactors have already been brought into permanent shutdowns, followed by decommissioning activities or preparation of decommissioning projects. On the occasion of the recent start of the decommissioning work at the Tokai Power Station, an overview on progress status in shutdown graphite-moderated power plants in several countries is given. In this report are described strategic aspects and some specific dismantling and waste management methods to be notified in individual decommissioning projects, as in the following. A few UK Magnox power stations have been in preparation for 'Safestore Construction', which will be reserved for more than 100 years after shutdown. The UKAEA's WAGR has been long undertaken as one of the big EC's reactor decommissioning projects, with extensive R and D work carried out for immediate dismantling of the graphite-moderated reactor. The recent successful progresses have revealed safe and commercial-scale dismantling procedures and technologies, which may facilitate an early dismantling shutdown nuclear facilities. The French GCR plants have been in plant-by-plant preparation for safestore for 30-40 years. The Spanish Vandellos-1 and Italian Latina plants are also under decommissioning operations similarly as in UK and France. All experimental and prototype high temperature reactor plants in Germany and USA had already been under decommissioning processes, with various safestore conditions depending on the specific project circumstances. The German AVR is being prepared for step-by-step dismantling the reactor structure. The Beloyarsk NPP based on ex-Soviet Union graphite reactor concept is still in preparatory phase in

  10. Radiation-induced deformation of graphite masonry of the Bilibin NPP reactor EhGP-6

    International Nuclear Information System (INIS)

    Results of measuring hole diameters in graphite masonry bricks of EhGP-6 reactors of the Bilibin NPP, depending on fast neutron fluence, are presented. Anomaly of deformation of front projections of bricks, related with their position in the core, was revealed and explained. Moment theory of nonaxisymmetric deformation of cylindrical cover was used to evaluate possible consequences of reduction of hole diameters in reactor masonry. 6 refs., 6 figs

  11. Nuclear graphite development, operational problems, and resolution of these problems at the Hanford production reactors

    International Nuclear Information System (INIS)

    This paper chronicles the history of the Hanford Production Reactor, from the initial design considerations for B, D, and F Reactors through the selection of the agreed method for safe disposal of the decommissioned reactors. The operational problems that challenged the operations and support staff of each new generation of production reactors, the engineering actions an operational changes that alleviated or resolved the immediate problems, the changes in reactor design and design-bases for the next generation of production reactors, and the changes in manufacturing variables that resulted in new ''improved'' grades of nuclear graphites for use in the moderators of the Hanford Production Reactors are reviewed in the context of the existing knowledge-base and the mission-driven priorities on the time. 14 refs, 6 figs, 3 tabs

  12. Present status of research reactor decommissioning programme in Indonesia

    International Nuclear Information System (INIS)

    At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)

  13. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    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)

  14. Physical security at research reactors

    International Nuclear Information System (INIS)

    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

  15. Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

    International Nuclear Information System (INIS)

    A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment

  16. Graphite materials testing in the ATR for lifetime management of Magnox reactors

    International Nuclear Information System (INIS)

    A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment. (author)

  17. Method to assess the radionuclide inventory of irradiated graphite waste from gas-cooled reactors

    International Nuclear Information System (INIS)

    About 17,000 tons of irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France Limited company (EDF), an energetic utility (http://www.edf.com/the-edf-group-42667.html) in France. Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. This huge discrepancy cannot be avoided and can be easily explained by Pierre Gy's theory of sampling of finely divided materials. The assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible whilst guaranteeing an upper margin corresponding to a 2.5 % risk of under-assessment. This method closely reflects the reality of the main phenomenon at radionuclide origin in a reactor, while also incorporating the secondary effects that can alter this result such as radionuclide or precursor release during reactor operation. (author)

  18. Magnesium hydroxide – expanded graphite composite pellets for a packed bed reactor chemical heat pump

    International Nuclear Information System (INIS)

    The chemical heat pump is a promising technology for the recovery of waste heat from industrial processes or cogeneration systems. It can be used for storing the surplus heat during low demand periods and release it for shaving the peaks of heat demand, with a benefit for the overall system efficiency. In this work, a packed bed reactor chemical heat pump based on the dehydration and hydration of magnesium hydroxide has been investigated. Due to its high thermal conductivity, expanded graphite was mixed with magnesium hydroxide to enhance heat transfer. The composite material, named EM, was developed and tested experimentally in order to understand the effects of expanded graphite on the chemical reactions occurring in the packed bed reactor. -- Highlights: • An expanded graphite/Mg(OH)2 composite was developed for a packed bed reactor chemical heat pump. • The expanded graphite/Mg(OH)2 composite (EM) was compressed in figure of pellets. • Higher reaction rates were observed for the dehydration and hydration of EM pellets. • EM pellets showed better performance in terms of heat storage and heat output. • EM pellets were able to withstand repetitive cyclic reactions without significant failures

  19. An automatic regulating control system for a graphite moderated reactor using digital techniques

    International Nuclear Information System (INIS)

    The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

  20. Management of research reactor ageing

    International Nuclear Information System (INIS)

    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

  1. Research reactors and alternative devices for research

    International Nuclear Information System (INIS)

    This report includes papers on research reactors and alternatives to the research reactors - radioisotopic neutron sources, cyclotrons, D-T neutron generators and small accelerators, used for radioisotope production, neutron activation analysis, material science, applied and basic research using neutron beams. A separate abstract was prepared for each of the 7 papers

  2. Proceedings of RIKEN BNL Research Center Workshop: Brookhaven Summer Program on Nucleon Spin Physics

    Energy Technology Data Exchange (ETDEWEB)

    Aschenauer, A.; Qiu, Jianwei; Vogelsang, W.; Yuan, F.

    2011-08-02

    Understanding the structure of the nucleon is of fundamental importance in sub-atomic physics. Already the experimental studies on the electro-magnetic form factors in the 1950s showed that the nucleon has a nontrivial internal structure, and the deep inelastic scattering experiments in the 1970s revealed the partonic substructure of the nucleon. Modern research focuses in particular on the spin and the gluonic structure of the nucleon. Experiments using deep inelastic scattering or polarized p-p collisions are carried out in the US at the CEBAF and RHIC facilities, respectively, and there are other experimental facilities around the world. More than twenty years ago, the European Muon Collaboration published their first experimental results on the proton spin structure as revealed in polarized deep inelastic lepton-nucleon scattering, and concluded that quarks contribute very little to the proton's spin. With additional experimental and theoretical investigations and progress in the following years, it is now established that, contrary to naive quark model expectations, quarks and anti-quarks carry only about 30% of the total spin of the proton. Twenty years later, the discovery from the polarized hadron collider at RHIC was equally surprising. For the phase space probed by existing RHIC experiments, gluons do not seem to contribute any to the proton's spin. To find out what carries the remaining part of proton's spin is a key focus in current hadronic physics and also a major driving force for the new generation of spin experiments at RHIC and Jefferson Lab and at a future Electron Ion Collider. It is therefore very important and timely to organize a series of annual spin physics meetings to summarize the status of proton spin physics, to focus the effort, and to layout the future perspectives. This summer program on 'Nucleon Spin Physics' held at Brookhaven National Laboratory (BNL) on July 14-27, 2010 [http://www.bnl.gov/spnsp/] is the

  3. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  4. Fusion reactor research

    International Nuclear Information System (INIS)

    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

  5. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

    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)

  6. Present and future directions of atomic physics research with multiply-charged ions at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Atomic physics research with multiply-charged ions is now in progress using ion beams from the Brookhaven Double MP-Tandem van de Graaff facility. In the near future, experiments will start using ions produced by photons from the National Synchrotron Light Source (NSLS). Examples of typical experiments are discussed to illustrate the comprehensive nature of these facilities. Plans for future expansion by addition of a CRYEBIS type ion source coupled to a heavy-ion storage ring for use in crossed-beam experiments at the NSLS are discussed. 18 refs., 8 figs

  7. Graphite structures

    International Nuclear Information System (INIS)

    Design criteria and operating experience of nuclear reactor graphite structures are reviewed on the basis of publications available since as early as 1954. Typical spatial distributions of temperature and neutron fluence are presented. Design features are discussed of graphite cores of the First-in-the-world NPP, the IR reactor, the Beloyarsk NPP, the RBMK reactor (all lwgr type), as well as of some GCR and HTGR type reactors now in operation in the U.S.A., Great Britain and the FRG. The improvement of graphite structure designs and their operating regimes take place through the development of means to combat the accumulation of latent energy (the Vigner energy), compensation of graphite dimensional changes, lowering gas leakages in gas-cooled reactors and preventing graphite structure oxidation. The data given on radiation-induced dimensional changes and stress states of the structures at temperatures up to 800 deg C testify to great potential possibilities of graphite application in reactor cores

  8. Replacement research reactor for Australia

    International Nuclear Information System (INIS)

    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)

  9. The WWR-SM-20 research reactor

    International Nuclear Information System (INIS)

    In this paper the design features and experimental capabilities of the WWR-SM-20 research reactor are described. The reactor uses fuel assemblies consisting of six coaxial fuel tubes with a square cross-section. IRT-3M fuel assemblies can be used with both 90% enriched and 36% enriched uranium. The main characteristics of the IRT-3M fuel assemblies are given, as are the technical and physical parameters of the WWR-SM-20 reactor. The core can hold up to ten ampoule-type channels with a diameter of up to 68 mm. For irradiation purposes, up to 22 26-mm-diameter channels in the fuel assemblies, and up to 48 42-mm-diameter channels in the beryllium blocks of the reflector can be used. In the graphite blanket between the horizontal channels, channels with a diameter of up to 130 mm can be used. The thermal neutron flux density has a maximum value of 1.5 X 1018 m-2 · s-1 in the core and 2.3 X 1018 m-2 · s-1 in the reflector, and the fast neutron flux density (cE > 0.821 MeV) a maximum of 1.9 X 1018 m-2 · s-1. A number of design features have been incorporated in the WWR-SM-20 reactor to make it effectively safe

  10. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    International Nuclear Information System (INIS)

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  11. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  12. Conditioning of graphite bricks from dismantled gas cooled reactors for disposal

    International Nuclear Information System (INIS)

    Dismantling of gas-cooled reactors to decommissioning involves tens of thousands of low-level radioactive graphite bricks containing less than 400 GBq of tritium per metric ton, less than 20 GBq.t-1 of 14C and less than 2 GBq of 36C1. The long half-life of the last two nuclides may require long-term conditioning. Core impregnation is one way to minimize leaching under these conditions. The authors show that core impregnation is possible using extremely durable pitches or bitumens, that inspection is possible by porosity monitoring and X-ray examination, and that satisfactory leach test results have been obtained on actual graphite samples from a reactor operated for 20 years. The very simple technology required for industrial implementation is discussed

  13. Advances in reactor safety research

    International Nuclear Information System (INIS)

    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.)

  14. Research reactor decommissioning

    International Nuclear Information System (INIS)

    Full text: Of the ∼ 800 research reactors constructed worldwide to date, ∼50% have been shut down and are at various stages of decommissioning. Many reached the end of their design lives or were shut down due to strategic, economic or regulatory considerations. 27% of those in operation are over 40 years old and will need to be decommissioned soon. Decommissioning normally takes the facility permanently out of service and subjects it to progressive hazard reduction, dismantling and decontamination in a safe, secure economically viable way, using best practicable means to meet the best practicable environmental option, such that the risks and doses to workers and the general public are maintained as low as reasonably practicable. Whilst most decommissioning techniques are well established there are still some challenging and important issues that need resolution. Perhaps the most challenging issue is radioactive waste management and storage. It is vitally important that all local and national waste classification, transportation, storage and end point requirements are known, as the adopted strategy will be heavily influenced by these factors. Other equally important but softer issues include the requirement for early decommissioning plans, adequate funding/cost estimates and the involvement of all relevant stakeholders. A comprehensive decommissioning plan should be produced up front that encompasses an early radiological characterisation survey of the facility/site. An appropriate funding mechanism needs to be assured. Whilst regular revisions of the decommissioning cost study should help to determine required funds, it is important to validate these cost estimates by benchmarking other decommissioning projects and accumulated experience. The use of appropriate 'stakeholder dialogue' methods by the facility operator to inform and communicate with all interested parties, such as government and non-government organisations, regulators, trades unions, anti

  15. Research Reactors Coalitions

    International Nuclear Information System (INIS)

    When considering the potential role of an existing RR or possibly the construction of a new RR, it is clear that a nuclear science and technology programme (including nuclear power) could benefit provided the RR is safely and competently managed, well utilised and adequately funded. Based on MSs experience, a domestic RR may not be required to develop a nuclear power programme, provided the decision takes advantage of foreign expertise, including access to foreign RRs facilities and RRs regional/international networks. If a country decides to gain access to a foreign research reactor, it may need considering the potential risk of change in the political relationship with the host country that could compromise the achievement of its national relevant objectives. This risk may be offset by availability of many options within one or more regional/international RRs networks and coalitions. Examples include the use of existing RRs in vendor, non-vendor countries and, in some cases non-nuclear power countries, to develop human resources in support of the introduction of nuclear power elsewhere. International RR networking trends are most evident with high flux, higher capability, and more complex fuel and material testing RRs being shared through international partnerships. However, networks involving low-medium power RRs for education and training purposes are also gaining a more prominent role to support nuclear capacity building in newcomer MSs. Networking through the internet seems also to be a promising way to support, as complementary offer to direct access to RRs facilities, MSs nuclear capacity building objectives (e.g. the IRL project)

  16. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.)

  17. Overview of strength, crack propagation and fracture of nuclear reactor moderator graphite

    Energy Technology Data Exchange (ETDEWEB)

    Moskovic, R., E-mail: robert.moskovic@magnoxsites.com [Magnox Limited, Oldbury Technical Centre, Oldbury Naite, South Gloucestershire BS35 1RQ (United Kingdom); Heard, P.J. [Interface Analysis Centre, University of Bristol, Bristol BS2 8BS (United Kingdom); Flewitt, P.E.J. [Magnox Limited, Oldbury Technical Centre, Oldbury Naite, South Gloucestershire BS35 1RQ (United Kingdom); Interface Analysis Centre, University of Bristol, Bristol BS2 8BS (United Kingdom); H.H. Wills Laboratory, Department of Physics, University of Bristol, Bristol BS8 1TL (United Kingdom); Wootton, M.R. [Magnox Limited, Oldbury Technical Centre, Oldbury Naite, South Gloucestershire BS35 1RQ (United Kingdom)

    2013-10-15

    Highlights: • Fracture behaviour. • Cracking initiation and growth. • Different loadings configurations. • Fracture mechanisms. -- Abstract: Nuclear reactor moderator graphite is an aggregate of needle coke filler particles within a matrix of fine coke flour particles mixed with pitch binder. Following extrusion in green condition, impregnation with liquid pitch binder and graphitisation, a polygranular aggregate with orthotropic properties is produced. Its mechanical properties under several different loading conditions and associated cracking behaviour were examined to establish crack initiation and propagation behaviour. Both virgin and radiolytically oxidised material were examined using optical and electron optical microscopy, focused ion beam microscope and digital image correlation. The appearance of force vs. displacement curves varied with type of loading. Mostly linear elastic traces occurred in uniaxial tensile and flexural tests. Large departures from linear elastic behaviour were observed in standard uniaxial and diametral compression testing. Digital image correlation has shown that the initiation of cracking involves formation of a process zone which grows to a critical size of approximately 3–5 mm before a macro-crack is initiated. Cracks straddle a torturous path which zigzags between the filler particles through the matrix consistent with crack propagation along the filler matrix interface. This paper provides an overview of strength, crack propagation and fracture of nuclear reactor moderator graphite. It reviews the physical processes and mathematical approaches that have been adopted to describe the behaviour of brittle materials and then considers if they apply to reactor core graphites.

  18. Loss of Coolant Accident Analysis for Israel Research Reactor

    International Nuclear Information System (INIS)

    One of the main objectives of reactor safety systems is to keep the reactor core in condition that does not permit any release of radioactivity into the environment. In order to ensure this, the reactor must have sufficient safety margins during all possible operational and accident conditions. This paper focuses on the analysis of loss of coolant accident (LOCA), which is one of the most severe scenarios among other hypothetical events such as reactivity induced accidents, loss of flow accident, etc. The analysis was carried out for the Israel Research Reactor 1 (IRR-1), which is a 5MW swimming pool type research reactor. The IRR-1 core consists of MTR highlyenriched uranium (HEU) fuel type, and is reflected by Graphite elements. During normal operation, the reactor core is cooled by downward forced flow of light water circulated by a primary cooling circuit pump. But during shutdown stage, the reactor core is cooled by upward natural convection flow through a safety flapper valve. There could be several primary causes to initiate a LOCA in research reactors, such as breaks in the piping system, ruptures of the beam tubes, and concrete wall failures of the reactor pool. Although probability of large break accident in research reactors is very low, once the accident occurs, it may cause major core damages, so it must be considered

  19. An explication of the graphite structural design code of core support components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    The integrity evaluation of the core support graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core support components. In the application of this design code, it is necessary to clear the basic concept to evaluate the core support components of HTTR. Therefore, considering the detailed design of core support graphite structures of HTTR, this report explicates the design code in detail about the concepts of stress limit, fatigue limit, other special stress limit and integrity evaluation of oxidized graphite components etc.. (author)

  20. The prospects of using carbon-graphite materials as construction elements of the microwave plasma diagnostic in a fusion reactor

    International Nuclear Information System (INIS)

    The comparative study of working characteristics of some graphite and metallic components of SHF diagnostics, such as several kinds of antennae, plane reflectors and cylindrical waveguides are presented. To simulate the consequence of a plasma disruption in large tokamaks on graphite-made antenna, we tested two antennae (MPG and ARV kinds of graphite) on the influence of power plasma flow: W≥1.10 MJ/m2 with the time duration Δt∼0.1 ms. All results show that the prospects to use graphite elements in a reactor-scale fusion device are rather promising. ((orig.))

  1. Enrichment reduction for research reactors

    International Nuclear Information System (INIS)

    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.)

  2. Ageing management for research reactors

    International Nuclear Information System (INIS)

    During the past several years, ageing of research reactor facilities continues to be an important safety issue. Despite the efforts exerted by operating organizations and regulatory authorities worldwide to address this issue, the need for an improved strategy as well as the need for establishing and implementing a systematic approach to ageing management at research reactors was identified. This paper discusses, on the basis of the IAEA Safety Standards, the effect of ageing on the safety of research reactors and presents a proactive strategy for ageing management. A systematic approach for ageing management is developed and presented together with its key elements, along with practical examples for their application. (author)

  3. The Design of Control-Rod Drives for Large Graphite-Moderated Reactors

    International Nuclear Information System (INIS)

    Because graphite-moderated tube-type power or desalinisation reactors are more economical in the larger ratings, control-rod drives may require strokes in the 20 to 60 ft range. Speed-of-insertion requirements may vary by a factor of 300 to 1 between the low-speed normal control requirements and the high-speed emergency shutdown requirements. Internal rod cooling is often required in addition to the prevention of reactor atmosphere leakage where the control rod penetrates the .reactor envelope. These requirements in addition to those of rod deceleration, shielding, space limitations, stored or emergency energy sources, maintenance provisions and overall drive-system cost increase the design problems associated with control rods for this type of reactor. Several unique control and/or shutdown rod drives have been designed for horizontal and vertical operation in large graphite-moderated power and study reactors. These designs include (1) air-operated shutdown rods with high insertion speeds, (2) hydraulic motor-driven, chain-type shutdown control rods with short storage sections and a compact drive; and (3) hydraulic cylinder-operated, force-multiplication shutdown control rods. Each of these drives compromises the requirements listed above to some extent; however, operable drives have been designed and tested. (author)

  4. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.)

  5. IAEA safeguards at research reactors

    International Nuclear Information System (INIS)

    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)

  6. Review: BNL graphite blanket design concepts

    International Nuclear Information System (INIS)

    A review of the Brookhaven National Laboratory (BNL) minimum activity graphite blanket designs is made. Three designs are identified and discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a thick graphite screen (typically 30 cm or greater, depending on type as well as application-experimental power reactor or commercial reactor). Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy. This energy is then either radiated to a secondary blanket with coolant tubes, as in types A and B, or is removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the structural material of the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude by the graphite screen, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma, whatever the degree of radiation damage

  7. Light water reactor safety research

    International Nuclear Information System (INIS)

    As the technology of light water reactors (LWR) was being commercialized, the German Federal Government funded the reactor safety research program, which was conducted by national research centers, universities, and industry, and which led to the establishment, in early 1972, of the Nuclear Safety Project in Karlsruhe. In the seventies, the PNS project mainly studied the loss-of-coolant accident. Numerous experiments were run and computer codes developed for this purpose. In the eighties, the Karlsruhe Nuclear Research Center contributed to the German Risk Study, investigating especially core meltdown accidents under the impact of the events at Three Mile Island-2 and Chernobyl-4. Safety research in the nineties is concentrated on the requirements of future reactor generations, such as the European Pressurized Water Reactor (EPR) or potential approaches which, at the present time, are discernible only as tentative theoretical designs. (orig.)

  8. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    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.)

  9. Research on the phenomenon of graphitization. Crystallographic study - Study of bromine sorption

    International Nuclear Information System (INIS)

    This research thesis reports the study of the mechanism of graphitization of carbon by using X-ray diffraction analysis and the physical and chemical study of lamellar reactions between carbon and bromine. The author first presents generalities and results of preliminary studies (meaning of graphitization, presentation of the various carbon groups and classes), and then reports the study of the graphitization of compact carbons (soft carbons). More precisely, he reports the crystallographic study of partially graphitized carbons: methods and principles, experimental results and their analysis, discussion of the graphitization mechanism. In the next part, the author reports the study of bromine sorption on carbons: experimental method, isotherms of a natural graphite and of a graphitized carbon, structure of carbon-bromine complexes, isotherms of graphitizable carbons and of all other carbons, distribution of bromine layers in partially graphitized carbons, bromine sorption and Fermi level

  10. Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

    International Nuclear Information System (INIS)

    Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed

  11. Data acquisition system for monitoring the graphite reflector temperatures in Cirus reactor

    International Nuclear Information System (INIS)

    A data acquisition system has been designed and developed at MS and ESS, BARC and is being used for monitoring Graphite reflector temperature of CIRUS reactor, BARC for last one year. It acquires the temperatures of various parts of (a) the graphite reflector, (b) the cast iron shield and (c) the air inlet and outlet, and also signals from various transmitters. The system also monitors these parameters every 500 ms for alarm conditions and stores the data for offline analysis. The data acquisition system consists of (1) a single board computer (SBC) based stand alone data acquisition system (SDAS) and (2) a personal computer (PC) based control and graphic user interface (GUI) system and they are linked by a serial interface. Since it is a critical system in terms of accuracy and reliability, various measures have been taken in the design and fabrication of all the subsystems. With this system now it is possible to monitor effect of various parameters on graphite reflector temperature of CIRUS reactor. (author)

  12. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  13. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    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)

  14. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  15. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    International Nuclear Information System (INIS)

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

  16. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO2-cooled reactors and for the decontamination of irradiated graphite waste

    Science.gov (United States)

    Le Guillou, M.; Toulhoat, N.; Pipon, Y.; Moncoffre, N.; Khodja, H.

    2015-06-01

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO2-cooled nuclear fission reactors (called UNGG for "Uranium Naturel-Graphite-Gaz") to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D+ ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO2) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the total amount produced

  17. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  18. BROOKHAVEN: Booster commissioned

    International Nuclear Information System (INIS)

    The construction and first commissioning phase of the Booster synchrotron to inject into Brookhaven's veteran Alternating Gradient Synchrotron (AGS) were completed last year. Scheduled to come into operation this year, the new Booster will extend the research capabilities AGS, and with its ability to accelerate partially stripped heavy ions will play an essential role in the chain of accelerators serving the Relativistic Heavy Ion Collider (RHIC)

  19. Research on plasma core reactors

    Science.gov (United States)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  20. Brookhaven Highlights, January 1982-March 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kuper, J.B.H.; Rustad, M.C. (eds.)

    1983-01-01

    Research at Brookhaven National Laboratory is summarized. Major headings are high energy physics, physics and chemistry, life sciences, applied energy science, support activities and administration. (GHT)

  1. Application researches on doped graphites as plasma facing materials

    International Nuclear Information System (INIS)

    The recrystallized graphites were prepared with calcined coke filler, coal-tar pitch binder and additives of B4C, Si, Ti by the hot-pressing in order to investigate the thermo-mechanical properties and microstructure of the recrystallized graphites. Radiation damages of the recrystallized graphites by ion bombardment in HT-7 fusion environment were also studied. The experimental results showed that the recrystallized graphites doped 10%B4C exhibited higher bending strength, up to 104 Mpa, but lower thermal conductivity; on the contrary, the recrystallized graphites doped Si and Ti exhibited higher thermal conductivity but lower bending strength. Compared with pure graphite, the erosion-resistant of the recrystallized graphites was improved obviously. The thickness of the gradient SiC coating of BSTDG was thinned from 40-50 μm to 5μm, and the coating flaked off partially after irradiation in HT-7 tokamak. (authors)

  2. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

  3. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and

  4. Principle design and data of graphite components

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned

  5. Research reactor education and training

    International Nuclear Information System (INIS)

    CORYS T.E.S.S. and TECHNICATOME present in this document some of the questions that can be rightfully raised concerning education and training of nuclear facilities' staffs. At first, some answers illustrate the tackled generic topics: importance of training, building of a training program, usable tools for training purposes. Afterwards, this paper deals more specifically with research reactors as an actual training tool. The pedagogical advantages they can bring are illustrated through an example consisting in the description of the AZUR facility training capabilities followed by the detailed experiences CORYS T.E.S.S. and TECHNICATOME have both gathered and keeps on gaining using research reactors for training means. The experience shows that this incomparable training material is not necessarily reserved to huge companies or organisations' numerous personnel. It offers enough flexibility to be adapted to the specific needs of a thinner audience. Thus research reactor staffs can also take advantages of this training method. (author)

  6. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U3Si2) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×1014 cm-2s-1, and is controlled by a Hafnium control absorber rod and B4C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  7. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  8. Upgrading the NRU research reactor

    International Nuclear Information System (INIS)

    After a nearly two-year long detailed review, AECL Research decided that its NRU research reactor will complete its mission around the turn of the century. The company's original intentions for major refurbishment have been revised and upgrading work will now mainly comprise add-ons to existing systems - so that research projects and isotope production schedules can be met - and procedure modifications to ensure continued safe operation. (Author)

  9. Evaluation of graphite/steam interactions for ITER [International Thermonuclear Experimental Reactor] accident scenarios

    International Nuclear Information System (INIS)

    This paper presents the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during an accident involving coolant leakage into the plasma chamber of the International Thermonuclear Experimental Reactor (ITER). This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 and 1700 degree C. The effects of steam flow rate and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We used the relationships for GraphNOL N3M in thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 16 refs., 7 figs., 1 tab

  10. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Experimental Pile O, Graphite Low-energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor (WAGR)) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6000t of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale pile 1 and the Windscale AGR during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe while minimizing the associated costs and technical risks. These options include (but are not limited to): disposal as low-level waste (LLW), incineration, or encapsulation and disposal as intermediate-level waste (ILW). There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel, and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of comfort' have been sought from both the LLW and the ILW disposal organizations to support the development of these waste disposal

  11. Defuelling of the UTR-300 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.D.; Banford, H.M.; East, B.W. [Scottish Universities Research and Reactor Centre, Glasgow (United Kingdom)

    1997-07-01

    The UTR-300 reactor at the Scottish Universities Research and Reactor Centre was based on the original Argonaut design with two aluminium core tanks set in a graphite reflector each containing six fuel elements cooled and moderated by water flowing up through the tanks in a closed primary circuit. The fuel plates in the original 13-plate elements were uranium oxide-aluminium with a 22g loading of 90% {sup 235}U. After 7 years of operation at 100 kW (10 kW average), the maximum power was increased to 300 kW (30 kW average) and, in order to maintain the operational excess reactivity, it was necessary to add another plate to each element progressively over the years until they all contained 14 plates. These extra plates were uranium metal-aluminium with 24.5 g of 90% {sup 235}U. No further modification of the elements was possible and so, with reactivity steadily decreasing, and for a variety of other reasons, a decision was taken to cease operation in September 1995. This paper describes the procedures whereby the fuel was unloaded from the core into a UNIFETCH flask equipped with a specially designed rotating gamma ray shield and then transported on two separate loads to Dounreay for reprocessing. (author)

  12. Defuelling of the UTR-300 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.D.; Banford, H.M.; East, B.W. [Scottish Universities Research and Reactor Centre, Glasgow (United Kingdom)

    1997-07-01

    The UTR-300 reactor at the Scottish Universities Research and Reactor Centre was based on the original Argonaut design with two aluminium core tanks set in a graphite reflector each containing six fuel elements cooled and moderated by water flowing up through the tanks in a closed primary circuit. The fuel plates in the original 13-plate elements were uranium oxide-aluminium with a 22g loading of 90% {sup 235}U. After 7 years of operation at 100 kW (10 kW average), the maximum power was increased to 300 kW (30 kW average) and, in order to maintain the operational excess reactivity, it was necessary to add another plate to each element progressively over the years until they all contained 14 plates. These extra plates were uranium metal-aluminium with 24.5 g of 90% {sup 235}U. No further modification of the elements was possible and so, with reactivity steadily decreasing, and for a variety of other reasons, a decision was taken to cease operation in September 1995. This paper describes the procedures whereby the fuel was unloaded from the core into a UNIFETCH flask equipped with a specially designed rotating gamma ray shield and then transported on two separate loads to Dounreay for reprocessing. (author) 2 figs., 2 tabs., refs.

  13. Abstracts of reports of the scientific-technical seminar on pulsed research reactors

    International Nuclear Information System (INIS)

    Pulsed graphite reactor IGR was 40 on May 13, 1998. The reactor was constructed to study physical processes occurring in atomic reactors. The IGR creation was initiated by I.V. Kurchatov, the remarkable Soviet physicist and academician. During these years hundreds of atomics units were tested, more than two thousand tests were conducted using IGR reactor. There were also obtained experimental results acknowledged in the world and that will always be of present interest. In relation to this remarkable event, the scientific-technical seminar 'Pulsed Research Reactors'. Application, experimental Researches and results' was held from 9 to 11 of June, 1998

  14. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  15. Fast reactor research in Switzerland

    International Nuclear Information System (INIS)

    The small Swiss research program on fast reactors serves to further understanding of the role of LMFR for energy production and to convert radioactive waste to more environmentally benign forms. These activities are on the one hand the contribution to the comparison of advanced nuclear systems and bring on the other to our physical and engineers understanding. (author)

  16. How Big Science Came to Long Island: the Birth of Brookhaven Lab (429th Brookhaven Lecture)

    International Nuclear Information System (INIS)

    Robert P. Crease, historian for the U.S. Department of Energy's Brookhaven National Laboratory and Chair of the Philosophy Department at Stony Brook University, will give two talks on the Laboratory's history on October 31 and December 12. Crease's October 31 talk, titled 'How Big Science Came to Long Island: The Birth of Brookhaven Lab,' will cover the founding of the Laboratory soon after World War II as a peacetime facility to construct and maintain basic research facilities, such as nuclear reactors and particle accelerators, that were too large for single institutions to build and operate. He will discuss the key figures involved in starting the Laboratory, including Nobel laureates I.I. Rabi and Norman Ramsey, as well as Donald Dexter Van Slyke, one of the most renowned medical researchers in American history. Crease also will focus on the many problems that had to be overcome in creating the Laboratory and designing its first big machines, as well as the evolving relations of the Laboratory with the surrounding Long Island community and news media. Throughout his talk, Crease will tell fascinating stories about Brookhaven's scientists and their research.

  17. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

  18. Australia's replacement research reactor project

    International Nuclear Information System (INIS)

    HIFAR, a 10 MW tank type DIDO Class reactor has operated at the Lucas Heights Science and Technology Centre for 43 years. HIFAR and the 10 kW Argonaut reactor 'Moata' which is in the Care and Maintenance phase of decommissioning are Australia's only nuclear reactors. The initial purpose for HIFAR was for materials testing to support a nuclear power program. Changing community attitude through the 1970's and a Government decision not to proceed with a planned nuclear power reactor resulted in a reduction of materials testing activities and a greater emphasis being placed on neutron beam research and the production of radioisotopes, particularly for medical purposes. HIFAR is not fully capable of satisfying the expected increase in demand for medical radiopharmaceuticals beyond the next 5 years and the radial configuration of the beam tubes severely restricts the scope and efficiency of neutron beam research. In 1997 the Australian Government decided that a replacement research reactor should be built by the Australian Nuclear Science and Technology Organisation at Lucas Heights subject to favourable results of an Environmental Impact Study. The Ei identified no reasons on the grounds of safety, health, hazard or risk to prevent construction on the preferred site and it was decided in May 1999 that there were no environmental reasons why construction of the facility should not proceed. In recent years ANSTO has been reviewing the operation of HIFAR and observing international developments in reactor technology. Limitations in the flexibility and efficiency achievable in operation of a tank type reactor and the higher intrinsic safety sought in fundamental design resulted in an early decision that the replacement reactor must be a pool type having cleaner and higher intensity tangential neutron beams of wider energy range than those available from HIFAR. ANSTO has chosen to use it's own resources supported by specialised external knowledge and experience to identify

  19. Brookhaven highlights, October 1978-September 1979

    International Nuclear Information System (INIS)

    These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the National Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities

  20. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    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)

  1. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  2. ACCELERATED SITE TECHNOLOGY DEPLOYMENT COST AND PERFORMANCE REPORT COMPARABILITY OF ISOCS INSTRUMENT IN RADIONUCLIDE CHARACTERICATION AT BROOKHAVEN NATIONAL LABORATORY

    Energy Technology Data Exchange (ETDEWEB)

    KALB,P.; LUCKETT,L.; MILLER,K.; GOGOLAK,C.; MILIAN,L.

    2001-03-01

    This report describes a DOE Accelerated Site Technology Deployment project being conducted at Brookhaven National Laboratory to deploy innovative, radiological, in situ analytical techniques. The technologies are being deployed in support of efforts to characterize the Brookhaven Graphite Research Reactor (BGRR) facility, which is currently undergoing decontamination and decommissioning. This report focuses on the deployment of the Canberra Industries In Situ Object Counting System (ISOCS) and assesses its data comparability to baseline methods of sampling and laboratory analysis. The battery-operated, field deployable gamma spectrometer provides traditional spectra of counts as a function of gamma energy. The spectra are then converted to radionuclide concentration by applying innovative efficiency calculations using monte carlo statistical methods and pre-defined geometry templates in the analysis software. Measurement of gamma emitting radionuclides has been accomplished during characterization of several BGRR components including the Pile Fan Sump, Above Ground Ducts, contaminated cooling fans, and graphite pile internals. Cs-137 is the predominant gamma-emitting radionuclide identified, with smaller quantities of Co-60 and Am-241 detected. The Project used the Multi-Agency Radiation Survey and Site Investigation Manual guidance and the Data Quality Objectives process to provide direction for survey planning and data quality assessment. Analytical results have been used to calculate data quality indicators (DQI) for the ISOCS measurements. Among the DQIs assessed in the report are sensitivity, accuracy, precision, bias, and minimum detectable concentration. The assessment of the in situ data quality using the DQIs demonstrates that the ISOCS data quality can be comparable to definitive level laboratory analysis when the field instrument is supported by an appropriate Quality Assurance Project Plan. A discussion of the results obtained by ISOCS analysis of

  3. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG)

  4. Research reactor's role in Korea

    International Nuclear Information System (INIS)

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

  5. New research reactor for Australia

    International Nuclear Information System (INIS)

    HIFAR, Australia's major research reactor was commissioned in 1958 to test materials for an envisaged indigenous nuclear power industry. HIFAR is a Dido type reactor which is operated at 10 MW. With the decision in the early 1970's not to proceed to nuclear power, HIFAR was adapted to other uses and has served Australia well as a base for national nuclear competence; as a national facility for neutron scattering/beam research; as a source of radioisotopes for medical diagnosis and treatment; and as a source of export revenue from the neutron transmutation doping of silicon for the semiconductor industry. However, all of HIFAR's capabilities are becoming less than optimum by world and regional standards. Neutron beam facilities have been overtaken on the world scene by research reactors with increased neutron fluxes, cold sources, and improved beams and neutron guides. Radioisotope production capabilities, while adequate to meet Australia's needs, cannot be easily expanded to tap the growing world market in radiopharmaceuticals. Similarly, neutron transmutation doped silicon production, and export income from it, is limited at a time when the world market for this material is expanding. ANSTO has therefore embarked on a program to replace HIFAR with a new multi-purpose national facility for nuclear research and technology in the form of a reactor: a) for neutron beam research, - with a peak thermal flux of the order of three times higher than that from HIFAR, - with a cold neutron source, guides and beam hall, b) that has radioisotope production facilities that are as good as, or better than, those in HIFAR, c) that maximizes the potential for commercial irradiations to offset facility operating costs, d) that maximizes flexibility to accommodate variations in user requirements during the life of the facility. ANSTO's case for the new research reactor received significant support earlier this month with the tabling in Parliament of a report by the Australian Science

  6. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3rd, 1965. It can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. A survey of the research programmes performed at the TRIGA Mainz is given covering applications in basic research as well as applied science in nuclear chemistry and nuclear physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of scientists, teachers, students and technical personal. Important projects for the future of the TRIGA Mainz are the UCN (ultra cold neutrons) experiment, fast chemical separation, medical applications and the use of the NAA as well as the use of the reactor facility for the training of students in the fields of nuclear chemistry, nuclear physics and radiation protection. Taking into account the past and future operation schedule and the typically low burn-up of TRIGA fuel elements (∝4 g U-235/a), the reactor can be operated for at least the next decade taking into account the fresh fuel elements on stock and without changing spent fuels. (orig.)

  7. Fission production and actinides in the spent graphite of the reactor stacks of the Siberian chemical integrated plant

    International Nuclear Information System (INIS)

    The peculiarity of the accomplished studies consisted in the representative selection of the reactor graphite stacks samples and in the performance of the complex analysis of their radioactive contamination. The role of incidents in forming the graphite contamination by individual radionuclides is identified and their distribution in stacks is studied. The correlation between the content of various radionuclides is investigated. The schemes for evaluating their reserve in the graphite stack are plotted. The results on evaluating the radionuclides reserve in the graphite stack highly differ from the earlier forecasted ones. The fission products and actinides reserves are by 10 times lesser as it was fore coated earlier, which may essentially simplify dismantling and selection of utilization technologies

  8. Relocation of a Research Reactor

    International Nuclear Information System (INIS)

    The research reactor RA-8 is placed in Pilcaniyeu Technological Centre (PTC) in the province of Rio Negro, approximately 70 km east from San Carlos de Bariloche city. The first time the reactor went critical was in June 1997 and it is out of operation since March 1999. Due to the intention to relocate the reactor in Bariloche Atomic Centre (BAC) a study has been done in order to assess the technical and economical feasibility. The scope of this study covers the disassembly and transport from PTC to BAC. Relocation of the reactor will reduce costs, time and difficulties in the transport of personnel (operators, researcher, and students) to Pilcaniyeu allowing the performance of nuclear research as well as academic application. The RA-8 is basically a critical facility of enriched uranium with light water as moderator. It is a pool type reactor with low thermal power, maximum 100 W and nominal 10 W. The principal assembly and the associated systems are placed in the reactor hall: consisting of a core, tanks, block, nuclear and conventional instrumentation, moderator system and the neutron source system. Also there is a control room with computers for monitoring together with the safety and control systems. The core is inside two stainless steel concentric tanks communicated with each other that contain water during the operation. The technical feasibility consists in: radiological characterization of the facility, visual inspection of the systems, structures and components, dismantling engineering, mass estimation for disassembly, packing, transport and storage. Economical feasibility has been done in order to evaluate time and costs necessaries for the disassembly and transport from Pilcaniyeu to Bariloche. Regulatory aspects that must be fulfilled were considered in this study. Nothing detectable was found in water samples from pipes of the pumps' well. The systems, structures and components of the RA-8 present in general a good condition of preservation that would

  9. Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors

    International Nuclear Information System (INIS)

    This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides (14C and 36Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of 36Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% 36Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)

  10. Accident analysis in research reactors

    International Nuclear Information System (INIS)

    With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

  11. Argentina: Disposal aspects of RA-1 research reactor decommissioning waste

    International Nuclear Information System (INIS)

    The objective of the project is to analyze disposal aspects of waste from total dismantling of Argentinean research reactors, starting with the oldest one, 48 years old RA-1. In order to estimate decommissioning waste, data was collected from files, area monitoring, measurements, sampling to measure activity and composition, operational history and tracing of operational incidents. Measurements were complemented with neutron activation calculations. Decommissioning waste for RA-1 is estimated to be 71.5 metric tons, most of it concrete (57 tons), the rest being steels, lead and reflector graphite (4.8 tons). Due to their low specific activities, no disposal problems are foreseen in the case of metals and concrete. Disposal of aluminium, steel, lead and concrete is analyzed. On the contrary, as the country has no experience in managing graphite radioactive waste, work was concentrated on that material. Stored (Wigner) energy may exist in RA-1 graphite reflectors irradiated at room temperature. Evaluation of stored energy by calorimetric methods is proposed, and its annealing by inductive heating; HEPA filters should be used to deal with gaseous activity emissions, mainly Cl-36 and C-14. Galvanic corrosion, dust explosion, ignition and oxidation can be addressed and should not become disposal problems. Care must be taken with graphite dust generation and disposal, due to wetting and flotation problems. Lessons learned from the project are presented, and the benefits of sharing international experience are stressed. (author)

  12. The NASA Space Radiation Laboratory at Brookhaven National Laboratory: Preparation and delivery of ion beams for space radiation research

    International Nuclear Information System (INIS)

    The NASA Space Radiation Laboratory (NSRL) at Brookhaven National Laboratory (BNL) was commissioned in October 2002 and became operational in July 2003. The NSRL was constructed in collaboration with NASA for the purpose of performing space radiation research as part of the NASA space program. The NSRL can accept a wide variety of ions from BNL's Collider Accelerator Department (CAD) Booster accelerator. These ion beams are extracted from the accelerator with kinetic energies ranging from 0.05 to 3 GeV/nucleon. Many different beam conditions have been produced for experiments at NSRL. The facilities at BNL and the design of the NSRL facility permit a wide variety of beams to be produced with a great degree of flexibility in the delivery of ion beams to experiments. In this report we will describe the facility and its performance over the eight experimental run periods that have taken place since it became operational. We will also describe the current and future capabilities of the NSRL.

  13. The Australian Replacement Research Reactor

    Science.gov (United States)

    Kennedy, Shane; Robinson, Robert

    2004-03-01

    The 20-MW Australian Replacement Research Reactor represents possibly the greatest single research infrastructure investment in Australia's history. Construction of the facility has commenced, following award of the construction contract in July 2000, and the construction licence in April 2002. The project includes a large state-of-the-art liquid deuterium cold-neutron source and supermirror guides feeding a large modern guide hall, in which most of the instruments are placed. Alongside the guide hall, there is good provision of laboratory, office and space for support activities. While the facility has "space" for up to 18 instruments, the project has funding for an initial set of 8 instruments, which will be ready when the reactor is fully operational in July 2006. Instrument performance will be competitive with the best research-reactor facilities anywhere, and our goal is to be in the top 3 such facilities worldwide. Staff to lead the design effort and man these instruments have been hired on the international market from leading overseas facilities, and from within Australia, and 7 out of 8 instruments have been specified and costed. At present the instrumentation project carries 10contingency. An extensive dialogue has taken place with the domestic user community and our international peers, via various means including a series of workshops over the last 2 years covering all 8 instruments, emerging areas of application like biology and the earth sciences, and computing infrastructure for the instruments.

  14. Reactor safety research in Sweden

    International Nuclear Information System (INIS)

    Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

  15. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

  16. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  17. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    International Nuclear Information System (INIS)

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated

  18. Current status of the world's research reactors

    International Nuclear Information System (INIS)

    Data from the IAEA's Research Reactor Database (RRDB) provides information with respect to the status of the world's research reactors. Some summary data are given. Recent initiatives by the IAEA regarding communications and information flow with respect to research reactors are discussed. Future plans and perspectives are also introduced. (author)

  19. Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report. (LEW)

  20. Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

    International Nuclear Information System (INIS)

    Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report

  1. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  2. Influence of plasma impulse on graphite surface

    International Nuclear Information System (INIS)

    Graphite is candidate material for a defence barrier of fusion reactor so the knowledge of its futures is necessary in constructing of different fusion devices. Most part of researches were made with use of continuous plasma flow. This report presents the measurements of impulse plasma affect on the graphite surface. Plasma was generated on coaxial plasma accelerator CPA. Hydrogen and nitrogen were taken as plasma-forming gases. Graphite examples were tested in different regimes of CPA. Main characters of radiation process were measured. There were found the creation of graphite erosion out of the surface. Also there were found the injection of plasma forming gas into depth of graphite. 2 refs

  3. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  4. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    International Nuclear Information System (INIS)

    The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors

  5. Myrrha, new polyvalent research reactor

    International Nuclear Information System (INIS)

    Myrrha (Multi-purpose hybrid research reactor for high-tech applications) is the first prototype of sub-critical nuclear reactor driven by a particle accelerator (an ADS, accelerator-driven system) at semi-industrial scale (50-100 MW), a safe and easy-to-control technology. In an interview, the manager of this project recalls his curriculum, presents and comments the characteristics of Myrrha, outlines why these ADS are so interesting to produce radio-isotopes, comments the variety of countries and companies involved in this project, outlines the peculiarities of Myrrha in terms of safety and the main technological challenges (a mixing of lead and bismuth for the coolant, control of corrosion by oxygen, an improved reliability based on redundant design and fault tolerance, MOX as fuel). He also evokes competing technologies

  6. Decommissioning of Salaspils Research Reactor

    International Nuclear Information System (INIS)

    The Salaspils Research Reactor (SRR) is out of operation since July 1998 and the decommissioning of SRR was started in 1999 according to the decision of the Government of Latvia. The main decommissioning activities up to 2006 were connected with collecting and conditioning of historical radioactive wastes from different storages outside and inside of reactor hall. The total amount of dismantled materials was about 700 tons, more than 77 tons were conditioned in concrete containers for disposal in repository. The radioactive wastes management technology is discussed in the paper. It was found, that additional efforts must be spent for immobilization of radionuclides in cemented matrix to be comply with the wastes acceptance criteria. The investigations of mechanical stability of water-cement matrix are described and discussed in the paper

  7. Progress with the Australian replacement research reactor

    International Nuclear Information System (INIS)

    Construction of the new Australian Research Reactor, the replacement for the now 46 year old HIFAR research reactor, is approximately 80% completed. Construction of the reactor facility began in April 2002 at ANSTO's Lucas Heights site near Sydney and commissioning is still on track for late 2005. Some details of the progress of construction and licensing and an outline of ANSTO research related to the use of Zircaloy-4 in the core region and reflector vessel of the reactor are given. (author)

  8. Reactor operations informal montly report, May 1, 1995--May 31, 1995

    International Nuclear Information System (INIS)

    This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination

  9. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  10. Failure prediction of full-size reactor components from tensile specimen data on NBG-18 nuclear graphite

    Energy Technology Data Exchange (ETDEWEB)

    Hindley, Michael P., E-mail: makke@mweb.co.za [Pebble Bed Modular Reactor (Pty) Ltd., P.O. Box 9396, Centurion 0046 (South Africa); Blaine, Deborah C.; Groenwold, Albert A.; Becker, Thorsten H. [Department of Mechanical and Mechatronic Engineering, Stellenbosch University, Private Bag X1, Matieland 7602 (South Africa)

    2015-04-01

    Highlights: • Predicts failure on a full scale reactor component and compare it to experiments. • Shows the effect of volume on NBG-18 nuclear graphite failure prediction. • Provide independent verification of a previously published methodology. • Describe the influence of multiple locations of high stress on failure prediction. - Abstract: This paper concerns itself with predicting the failure of a full-size NBG-18 nuclear graphite reactor component based only on test data obtained from standard tensile test specimens. A full-size specimen structural test was developed to simulate the same failure conditions expected during a normal operation of the reactor in order to validate the failure prediction. The full-size specimen designed for this test is almost a hundred times larger than the tensile test specimen, has a completely different geometry and experiences a different loading condition to the standard tensile test specimen. Failure of the full-size component is predicted realistically, but conservatively.

  11. Operating experiences of the research reactors

    International Nuclear Information System (INIS)

    Nuclear research reactors are devices of wide importance, being used for different scientific research tasks, for testing and improving reactor systems and components, for the production of radioisotopes, for the purposes of defence, for staff training and for other purposes. There are three research reactors in Yugoslavia: RA, RB and TRIGA. Reactors RA and RB at the 'Boris Kidric' Institute of Nuclear Sciences are of heavy water type power being 6500 and 10 kW, and maximum thermal neutron flux of 1014 and 1011(n/cm2s), respectively. TRIGA reactor at the 'Jozef Stefan' Institute in Ljubljana is of 250 kW power and maximum thermal neutron flux of 1013(n/cm2s). Reactors RA and RB use soviet fuel in the form of uranium dioxide (80% enriched) and metallic uranium (2%). Besides, RB reactor operates with natural uranium too. TRIGA reactor uses american uranium fuel 70% and 20% enriched, uranium being mixed homogeneously with moderator (ZrH). Experiences in handling and controlling the fuel before irradiation in the reactor, in reactor and after it are numerous and valuable, involving either the commercial arrangements with foreign producers, or optimal burn up in reactor or fuel treatment after the reactor irradiation. Twenty years of operating experience of these reactors have great importance especially having in mind the number of trained staff. Maintenance of reactors systems and fluids in continuous operation is valuable experience from the point of view of water reactor utilization. The case of the RA reactor primary cycle cobalt decontamination and other events connected with nuclear and radiation security for all three reactors are also specially emphasized. Owing to our research reactors, numerous theoretical, numerical and experimental methods are developed for nuclear and other analyses and design of research and power reactors,as well as methods for control and protection of radiation. (author)

  12. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  13. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  14. Nuclear research reactors activities in INVAP

    International Nuclear Information System (INIS)

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented

  15. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  16. Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

    OpenAIRE

    Diez de la Obra, Carlos Javier; Stankovskiy, Alexey; Malambu, E.; Zerovnik, Gasper; Schillebeeckx, Peter; Van Den Eynde, Gert; Heyse, Jan; Cabellos de Francisco, Oscar Luis

    2013-01-01

    A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled re...

  17. Measurements of Shut-Down Margin in Water and Gas-Graphite Reactors; Principles and Value of the Method

    International Nuclear Information System (INIS)

    Recent accidents have drawn the attention of technologists to the problems of measuring reactor shut-down margin during testing and operation. This paper deals with principles for determining the shut-down margin in water and gas-graphite reactors and the corresponding measurement techniques. These techniques are still not fully worked out, and the methods of calculating and preparing data for the two types of reactor considered here are rather difficult and costly. The paper sums up and outlines the problems that need further study. (author)

  18. Technical aspects of boron neutron capture therapy at the BNL Medical Research Reactor

    International Nuclear Information System (INIS)

    The Brookhaven Medical Research Reactor, BMRR, is a 3 MW heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies. Early BNL work in Boron Neutron Capture Therapy (BNCT) used a beam of thermal neutrons for experimental treatment of brain tumors. Research elsewhere and at BNL indicated that higher energy neutrons would be required to treat deep seated brain tumors. Epithermal neutrons would be thermalized as they penetrated the brain and peak thermal neutron flux densities would occur at the depth of brain tumors. One of the two BMRR thermal port shutters was modified in 1988 to include plates of aluminum and aluminum oxide to provide an epithermal port. Lithium carbonate in polyethylene was added in 1991 around the bismuth port to reduce the neutron flux density coming from outside the port. To enhance the epithermal neutron flux density, the two vertical thimbles A-3 (core edge) and E-3 (in core) were replaced with fuel elements. There are now four fuel elements of 190 grams each and 28 fuel elements of 140 grams each for a total of 4.68 kg of 235U in the core. The authors have proposed replacing the epithermal shutter with a fission converter plate shutter. It is estimated that the new shutter would increase the epithermal neutron flux density by a factor of seven and the epithermal/fast neutron ratio by a factor of two. The modifications made to the BMRR in the past few years permit BNCT for brain tumors without the need to reflect scalp and bone flaps. Radiation workers are monitored via a TLD badge and a self-reading dosimeter during each experiment. An early concern was raised about whether workers would be subject to a significant dose rate from working with patients who have been irradiated. The gamma ray doses for the representative key personnel involved in the care of the first 12 patients receiving BNCT are listed. These workers did not receive unusually high exposures

  19. Utilisation of the Research Reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    The TRIGA Mark II reactor of the University of Mainz can be operated in the steady state mode with thermal powers up to a maximum of 100 kW and in the pulse mode with a maximum peak power of 250 MW. So far, more than 17 000 pulses have been performed. For irradiations the TRIGA Mainz has a central experimental tube, three pneumatic transfer systems and a rotary specimen rack. In addition, the TRIGA Mainz includes four horizontal beam ports and a graphite thermal column which provides a source of well-thermalised neutrons. A broad spectrum of commercial applications, scientific research and training can be executed. For education and training various courses in nuclear and radiochemistry, radiation protection, reactor operation and physics are held for scientists, advanced students, teachers, engineers and technicians. Isotope production and Neutron Activation Analysis (NAA) are applied in in-core positions for different applications. NAA in Mainz is focused to determine trace elements in different materials such as in archaeometry, forensics, biology and technical materials including semiconductors for photovoltaics. The beam ports and the thermal column are used for commercial as well as for special basic and applied research in medicine, biology, chemistry and physics. Experiments are in preparation to determine the fundamental neutron properties with very high precision using ultra cold neutrons (UCN) produced at the tangential beam port. A second source is under development at the radial piercing beam port. Another experiment under development is the determination of ground-state properties of radioactive nuclei with very high precision using a penning trap and collinear laser spectroscopy. For many years fast chemical separation procedures combining a gas-jet transport system installed in one beam tube with either continuous or discontinuous chemical separation are carried out. In addition the thermal column of the reactor is also used for medical and

  20. Reliability studies in research reactors

    International Nuclear Information System (INIS)

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  1. Brookhaven highlights, 1986-1987

    International Nuclear Information System (INIS)

    The highlights of research conducted between October 1985 and September 1987 at Brookhaven National Laboratory are reviewed in this publication. Also covered are the administrative and financial status of the laboratory and a brief mention of meetings held and honors received. (FI)

  2. Dr. Praveen Chaudhari named director of Brookhaven National Laboratory

    CERN Multimedia

    2003-01-01

    "Brookhaven Science Associates announced today the selection of Dr. Praveen Chaudhari as Director of the U.S. Department of Energy's Brookhaven National Laboratory. Dr. Chaudhari, who will begin his new duties on April 1, joins Brookhaven Lab after 36 years of distinguished service at IBM as a scientist and senior manager of research" (1 page).

  3. A Multi-Phased Sampling Effort to Characterize a University TRIGA Research Reactor

    International Nuclear Information System (INIS)

    A radiological characterization project was conducted at the University of Illinois (University) TRIGA research nuclear reactor in July 2005 as part of the long-term facility decommissioning project. The characterization effort included multiple survey and sampling techniques designed to assess both contamination of the reactor building and equipment and activation of reactor components and the reactor bio-shield. Radiation measurements included alpha and beta surface contamination measurements, gamma dose rate measurements, and gross gamma radiation measurements. Modeling was conducted based on the field measurements to predict concentrations of activation products in reactor components that were not directly sampled. The sampling effort included collecting removable contamination swipes, concrete samples from the reactor room floor and bio-shield, soil samples from below and around the perimeter of the reactor building, graphite samples from graphite moderator, and metal samples from reactor components. Concrete samples were obtained using an innovative technology that allowed for quick sample collection and analysis. Concrete, soil, graphite, and metal samples were analyzed on-site using liquid scintillation counters and gamma spectroscopy. Additional samples were sent off-site for analysis. (authors)

  4. Overview on New Research Reactors in China

    International Nuclear Information System (INIS)

    In China, 2 research reactors are now under construction. Correspondingly, this paper consists of 2 parts. Part 1 will focus on China Advanced Research Reactor (CARR), the reactor characteristics, utilization, safety related systems and other main systems will be described in this part. Part 2 will focus on China Experiment Fast Reactor(CEFR), the general design and the safety features in particular will be illustrated in this part. (author)

  5. Reactor Safety Research Programs Quarterly Report January - March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  6. Use of research reactors for training and teaching nuclear sciences

    International Nuclear Information System (INIS)

    Training activities on reactors are organized by Cea on 2 specially dedicated reactors Ulysse (Saclay) and Siloette (Grenoble) and 2 research reactors Minerve (Cadarache) and Azur (Cadarache, facility managed by Technicatome). About 4000 students have been trained on Ulysse since its commissioning more than 40 years ago. The concept that led to the design of Ulysse was to build a true reactor dedicated to teaching and training activities and that was able to operate with great flexibility and under high conditions of safety, this reactor is inspired from the Argonaut-type reactor. The main specificities of Ulysse are: a nominal power of 100 kW, a maximal thermal neutron flux of 1.4 1012 n.cm-2.s-1, a 90 % enriched fuel, a graphite reflector, the use of water as coolant and moderator, and 6 cadmium plates as control rods. Ulysse allows students to get practical experience on a large range of topics: approach to criticality, effect of the starting neutron source, calibration of control rods, distribution of the neutron flux in the thermal column, temperature coefficient, radiation detectors, neutron activation analysis, and radioprotection. (A.C.)

  7. Basic research using the 250 kW research reactor of the Jozef Stefan Institute

    International Nuclear Information System (INIS)

    The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. The reactor therefore has a large prompt negative temperature coefficient of reactivity; the fuel also has a very high retention of radioactive fission products. The experimental facilities include a rotary specimen rack, a central in-core radiation thimble, a pneumatic transfer system and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x1013n/cm2 in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x1016n/cm2sec. All TRIGA reactors produce a core-average thermal neutron flux of about 107 n.v. per watt. Only with very large accelerators can such high fluxes be achieved. The types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine, in biology, archaeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. We can conclude that the 250 kW TRIGA reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

  8. Graphite components

    International Nuclear Information System (INIS)

    The reflector of the AVR reactor consists of needle coke graphite ARS/AMT made by the firm of Sigri GmbH. Although its orientation anisotropy is greater than those of materials developed in the last few years, the reflector graphite shows comparatively good dimensional stability in the conditions prevailing in the AVR reactor, as shown by tests on irradiation samples. It was also found that the strength increased. No damage was found during inspections in 1984 of the upper side and ceiling reflector. After a 20 year operating period, the graphite components of the AVR reactor should be in an excellent state. This will be possible to prove when samples are taken from the graphite reflector and examined in the context of a dismantling program. (orig.)

  9. Safe Operation of Research Reactors in Germany

    International Nuclear Information System (INIS)

    In Germany, experience was gained in the field of safe operation of research reactors during the last five decades. In this time, in total 46 research reactors were built and operated safely. Concerning the design, there is, or has been, a very broad range of different types of research reactors. The variety of facilities includes large pool or tank reactors with a thermal power of several tens of megawatt as well as small educational reactors with a negligible thermal power and critical assemblies. At present, 8 research reactors are still in operation. The other facilities are permanently shutdown, in decommissioning or have already been dismantled completely and released from regulatory control. In this paper, four selected facilities still being operated are presented as examples for safe operation of research reactors in Germany, including especially a description of the safety reviews and safety upgrades for the older facilities. (author)

  10. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  11. Application of research reactors for radiation education

    International Nuclear Information System (INIS)

    Nuclear research Reactors are, as well as being necessary for research purposes, indispensable educational tools for a country whose electric power resources are strongly dependent on nuclear energy. Both large and small research reactors are available, but small ones are highly useful from the viewpoint of radiation education. This paper oders a brief review of how small research reactors can, and must, be used for radiation education for high school students, college and graduate students, as well as for the public. (author)

  12. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    International Nuclear Information System (INIS)

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector

  13. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  14. Recovery of 14C from graphite moderator of gas-cooled reactor (GCR)

    International Nuclear Information System (INIS)

    The chemical exchange method of carbon isotopes between CO2 and carbamate was applied to the recovery of 14C from 1,600 t graphite moderator of a gas-cooled reactor (GCR), Tokai-1, and the dimensions of 14C-enrichment process were evaluated numerically. Applicability of two processes with different operation modes, continuous process and batch process, was discussed under the conditions that the concentration of 14CO2 in the stripped flow corresponding to 99% of feed CO2 is less than the environmental standard. For the continuous process using 2 mol/l diethylamine (DEA)-octane solution as a working fluid at -20degC and 0.2 MPa, the column dimensions were evaluated as 3.2 m in diameter and 5.7 m in height in the case of operating period of 20 yr. For the batch process using 4 mol/l DEA-octane solution, the column dimensions were comparable to those of continuos process, when the process was operated at the rate of 4 batch/month under the conditions of -20degC and 0.3 MPa. From these results, it is concluded that the CO2/carbamate exchange method is applicable to the recovery of 14C from irradiated graphite. However, the batch process has serious disadvantages, such as large energy consumption to maintain the top reservoir at low temperature and the generation of a large quantity of secondary wastes. At the present stage, the continuous process should be selected for the practical process design. (author)

  15. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

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

    International Nuclear Information System (INIS)

    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)

  17. Euratom research supporting reactor safety

    International Nuclear Information System (INIS)

    This paper focuses on the way the Fifth Framework Programme (FP5) of the European Atomic Energy Community (Euratom) (1998-2002) contributed to the overall effort supporting LWR safety and on the prospects of FP6 (2002-2006) to achieve the additional goal of contributing to the creation of the European Research Area (ERA). The objectives of the FP5 Community research are discussed, topic per topic, with reference to the relevant projects contributing to their achievement while emphasising the role and expectations of the end-users. 71 research projects have been carried out in FP5 with a total cost of 85.4 million Euro, out of which 43.5 million Euro are contributed by the EU. Economic aspects and needs for nuclear knowledge management throughout the EU-25 countries are briefly discussed as well as future Euratom research needs and nuclear stakeholders' interests. Implementation aspects and prospects of Euratom FP6 are discussed bearing in mind the challenging additional objective. Very ambitious S/T actions have been undertaken supporting reactor safety not only by their technical achievements but also by their structuring effect in the European research picture. (authors)

  18. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    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

  19. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

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

  20. Finite element modelling of the effect of temperature and neutron dose on the fracture behaviour of nuclear reactor graphite bricks

    International Nuclear Information System (INIS)

    Highlights: • Effects of irradiation on fracture behaviours of graphite bricks are analysed. • Two irradiation conditions chosen are irradiation temperature and neutron dose. • The crack initiates around the keyway fillet of the brick for every study. • Higher temperature and higher neutron dose accelerate crack initiation time. • Turnaround point of hoop strain indicates the crack initiation time. - Abstract: Graphite moderator bricks used within many UK gas-cooled nuclear reactors undergo harsh temperature and radiation gradients. They cause changes in material properties of graphite over extended periods of time. Consequently, models have been developed in order to understand and predict the complex stresses formed within the brick by these processes. In this paper the effect of irradiation temperature and neutron dose on the fracture characteristics, crack initiation and crack growth are investigated. A finite element (FE) mechanical constitutive model is implemented in combination with the damage model to simulate crack growth within the graphite brick. The damage model is based on a linear traction–separation cohesive model in conjunction with the extended finite element method for arbitrary crack initiation and propagation. Results obtained have showed that cracks initiate in the vicinity of the keyway fillet of the graphite brick and initiation time accelerates with higher temperatures and doses

  1. Finite element modelling of the effect of temperature and neutron dose on the fracture behaviour of nuclear reactor graphite bricks

    Energy Technology Data Exchange (ETDEWEB)

    Wadsworth, M.; Kyaw, S.T., E-mail: si.kyaw@nottingham.ac.uk; Sun, W.

    2014-12-15

    Highlights: • Effects of irradiation on fracture behaviours of graphite bricks are analysed. • Two irradiation conditions chosen are irradiation temperature and neutron dose. • The crack initiates around the keyway fillet of the brick for every study. • Higher temperature and higher neutron dose accelerate crack initiation time. • Turnaround point of hoop strain indicates the crack initiation time. - Abstract: Graphite moderator bricks used within many UK gas-cooled nuclear reactors undergo harsh temperature and radiation gradients. They cause changes in material properties of graphite over extended periods of time. Consequently, models have been developed in order to understand and predict the complex stresses formed within the brick by these processes. In this paper the effect of irradiation temperature and neutron dose on the fracture characteristics, crack initiation and crack growth are investigated. A finite element (FE) mechanical constitutive model is implemented in combination with the damage model to simulate crack growth within the graphite brick. The damage model is based on a linear traction–separation cohesive model in conjunction with the extended finite element method for arbitrary crack initiation and propagation. Results obtained have showed that cracks initiate in the vicinity of the keyway fillet of the graphite brick and initiation time accelerates with higher temperatures and doses.

  2. Fuel-failure detection system for Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Ayazuddin, S.K.; Hayat, Tariq; Qureshi, A.A.; Khan, H.A. [Pakistan Inst. of Nuclear Science and Technology, Nuclear Engineering Div., Islamabad (Pakistan)

    1997-12-01

    After the conversion and upgrading of Pakistan Research Reactor-1 (PARR-1), it was decided to install a fuel-failure detection system to confirm the performance and integrity of the new fuel elements. The fuel-failure detection is based on monitoring of delayed neutrons emitted from fission products leaking into the primary coolant loop from the fuel. For this purpose, two neutron detectors (BF{sub 3}) were replaced in the graphite moderator blocks that were installed at the outlet coolant pipe in the valve pit. The fuel-failure detection system was tested and calibrated at a miniature neutron source reactor (PARR-2) which provided the basis for alarm limits setting in the event of fuel failure. (author).

  3. Fuel-failure detection system for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    After the conversion and upgrading of Pakistan Research Reactor-1 (PARR-1), it was decided to install a fuel-failure detection system to confirm the performance and integrity of the new fuel elements. The fuel-failure detection is based on monitoring of delayed neutrons emitted from fission products leaking into the primary coolant loop from the fuel. For this purpose, two neutron detectors (BF3) were replaced in the graphite moderator blocks that were installed at the outlet coolant pipe in the valve pit. The fuel-failure detection system was tested and calibrated at a miniature neutron source reactor (PARR-2) which provided the basis for alarm limits setting in the event of fuel failure. (author)

  4. Light water reactor safety research project

    International Nuclear Information System (INIS)

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  5. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

    International Nuclear Information System (INIS)

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  6. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  7. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  8. REACT/THERMIX - a computer code to calculate graphite corrosion due to accidents in pebble-bed reactors

    International Nuclear Information System (INIS)

    his report presents the description of the computer code REACT/THERMIX, which was developed for calculations of the graphite corrosion phenomena and accident transients in gas cooled High Temperature Reactors (HTR) under air and/or water ingress accident conditions. The two-dimensional code is characterized by direct coupling of thermodynamic, fluiddynamic and chemical processes with a separate handling of heterogeneous chemical reactions. (orig.)

  9. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  10. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 3; Trench 5 at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed Final Status Survey (FSS) of the concrete duct from Trench 5 from Building 801 to the Stack. Sample results have been submitted as required to demonstrate that the cleanup goal of (le)15 mrem/yr above background to a resident in 50 years has been met. Four rounds of sampling, from pre-excavation to FSS, were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the U.S. Department of Energy (DOE) to perform independent verifications of decontamination and decommissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task for the HFBR Underground Utilities. ORISE, together with DOE, determined that a Type A verification of Trench 5 was appropriate based on recent verification results from Trenches 2, 3, and 4, and the minimal potential for residual radioactivity in the area. The removal of underground utilities is being performed in three stages to decommission the HFBR facility and support structures. Phase 3 of this project included the removal of at least 200 feet of 36-inch to 42-inch pipe from the west side to the south side of Building 801, and the 14-inch diameter Acid Waste Line that spanned from 801 to the Stack within Trench 5. Based on the pre-excavation sample results of the soil overburden the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the BNL FSP and identified comments for consideration (ORISE 2010). BNL prepared a revised FSP that resolved each ORISE comment adequately (BNL 2010a). ORISE referred to the revised HFBR Underground Utilities FSP FSS data to conduct the Type A verification

  11. Research nuclear reactor RA - Annual Report 2000

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 2000 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  12. Research nuclear reactor RA - Annual Report 1998

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 1998 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  13. Research nuclear reactor RA - Annual Report 1999

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. Research reactor RA Annual report for year 1998 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  14. Dose measurements and calculations in the epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

    International Nuclear Information System (INIS)

    The characteristics of the epithermal neutron beam at BMRR were measured, calculated, and reported by R.G. Fairchild. This beam has already been used for animal irradiations. The authors anticipate that it will be used for clinical trials. Thermal and epithermal neutron flux densities distributions, and dose rate distributions, as a function of depth were measured in a lucite dog-head phantom. Monte Carlo calculations were performed and compared with the measured values

  15. Boron neutron capture therapy of malignant brain tumors at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joel, D.D.; Coderre, J.A.; Chanana, A.D. [Brookhaven National Lab., Upton, NY (United States). Medical Dept.

    1996-12-31

    Boron neutron capture therapy (BNCT) is a bimodal form of radiation therapy for cancer. The first component of this treatment is the preferential localization of the stable isotope {sup 10}B in tumor cells by targeting with boronated compounds. The tumor and surrounding tissue is then irradiated with a neutron beam resulting in thermal neutron/{sup 10}B reactions ({sup 10}B(n,{alpha}){sup 7}Li) resulting in the production of localized high LET radiation from alpha and {sup 7}Li particles. These products of the neutron capture reaction are very damaging to cells, but of short range so that the majority of the ionizing energy released is microscopically confined to the vicinity of the boron-containing compound. In principal it should be possible with BNCT to selectively destroy small nests or even single cancer cells located within normal tissue. It follows that the major improvements in this form of radiation therapy are going to come largely from the development of boron compounds with greater tumor selectivity, although there will certainly be advances made in neutron beam quality as well as the possible development of alternative sources of neutron beams, particularly accelerator-based epithermal neutron beams.

  16. Usage of burnable poison on research reactors

    International Nuclear Information System (INIS)

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  17. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  18. The Berkeley TRIGA Mark III research reactor

    International Nuclear Information System (INIS)

    The Berkeley Research Reactor went critical on August 10, 1966, and achieved licensed operating power of 1000 kW shortly thereafter. Since then, the reactor has operated, by and large, trouble free on a one-shift basis. The major use of the reactor is in service irradiations, and many scientific programs are accommodated, both on and off campus. The principal off-campus user is the Lawrence Radiation Laboratory at Berkeley. The reactor is also an important instructional tool in the Nuclear Engineering Department reactor experiments laboratory course, and as a source of radioisotopes for two other laboratory courses given by the Department. Finally, the reactor is used in several research programs conducted within the Department, involving studies with neutron beams and in reactor kinetics

  19. Finite Element Based Stress Analysis of Graphite Component in High Temperature Gas Cooled Reactor Core Using Linear and Nonlinear Irradiation Creep Models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish; Majumdar, Saurindranath

    2015-01-01

    Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

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

    International Nuclear Information System (INIS)

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  1. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    Energy Technology Data Exchange (ETDEWEB)

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  2. Research reactor and its application in Thailand

    International Nuclear Information System (INIS)

    The first Thai Research Reactor (TRR-1) was established in 1961. TRR-1 had been operated with power of 1 MW from 1962 to 1975 and was shut down for modification during 1975 to 1977. The Thai Research Reactor1/modification1 (TRR-1/M1) is a multipurpose reactor with nominal power of 2 MW. Since 1977 TRR-1/M1 has been operated and utilized for various applications such as neutron activation analysis, radioisotope production, gem irradiation, neutron radiography and research works. To expand and promote the utilization of research reactor, the new 10 MW Research Reactor will be established in the Ongkarak Nuclear Research Center (ONRC) project and the project will be finished in the near future. (author)

  3. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  4. Energy-related perturbations of the northeast coastal zone: five years (1974-1979) of oceanographic research at Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, J.J.

    1980-03-01

    Since inception of oceanographic research at Brookhaven National Laboratory in 1974, over 75 cruises and 150 papers and reports have been completed. In comparison of shelf ecosystems at high, mid, and low latitudes, an understanding of the natural variability of US coastal waters has been derived. Annual carbon and nitrogen budgets suggest that the energy flow is diverted to a pelagic food web in summer-fall and a demersal food web in winter-spring within the Mid-Atlantic Bight. The impact of energy-related perturbations can now be assessed within the context of natural oscillation of the coastal food web.

  5. Graphite for high temperature gas-cooled reactors GT-MGR

    International Nuclear Information System (INIS)

    The properties of GR-1 graphite based on plentiful and economical raw material, i.e. unfired pitch coke for replaceable fuel blocks of the GT-MGR core, are presented. It is established that the optimal variant of this graphite meets the technical requirements, and its high linear thermal expansion coefficient makes it possible to expect adequate radiation dimensional stability. It is shown that with respect to a set of characteristics GR-1 graphite can be regarded as a candidate material

  6. Nondestructive testing on graphite structures for high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    The application of ultrasonic (for internal defects) and eddy current testing (for surface defects) were investigated on the structures of nuclear-grade IG-110 and PGX graphite for the HTTR. The equipment were developed in order to detect the specific configuration of graphite blocks and the testing conditions were defined as the practical testing methods. The established testing methods are being used for the acceptance tests of graphite structures in the HTTR. (author)

  7. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  8. Overview of research reactor operation within AECL

    International Nuclear Information System (INIS)

    This paper presents information on reactor operations within the Research Company of Atomic Energy of Canada (AECL) today relative to a few years ago, and speculates on future operations. In recent years, the need for Research Company reactors has diminished. This, combined with economic pressures, has led to the shutdown of some of the company's major reactors. However, compliance with the government agenda to privatize government companies in Canada, and a Research Company policy of business development, has led to some offsetting activities. The building of a pool-type 10 MWt MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor for isotope production will assist in the sale of the AECL isotopes marketing company. A Low Enriched Uranium (LEU) fuel fabrication facility and a Tritium Extraction Plant (TEP), both currently under construction, are needed in support of the NRU (National Research Universal) reactor and are in line with business development strategies. The research program demands on NRU stretch many years into the future and the strategies for achieving effective operation of this aging reactor, now 32 years old, are discussed. The repair of the leaking light-water reflector of the NRU reactor is highlighted. The isotope business requires that a second reactor be available for back-up production and the operation of the 42 year old NRX (National Research Experimental) reactor in its present 'hot standby' mode is believed to be unique in the world

  9. Design of a multipurpose research reactor

    International Nuclear Information System (INIS)

    The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

  10. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  11. Problems of Decommissioning Research Reactor IR-100

    International Nuclear Information System (INIS)

    The research reactor IR-100 with a thermal power of 200 kW is assigned to conduct science research and training activities in the fields of nuclear and molecular physics, radiation chemistry, radioactive isotope production, material, irradiation in neutron and gamma fields of devices and equipment, as well as for training of specialists for nuclear reactor operation

  12. Research Reactors - An analysis with a focus on non-proliferation and export control

    International Nuclear Information System (INIS)

    The Swedish Defence Research Agency, FOI, has under contract work financed by the Swedish Radiation Safety Authority, performed a study on research reactors. The principle of a research reactor and its characteristics and uses are described in this report. The potential use of research reactors for plutonium production for nuclear weapons is also described and the parameters of importance for optimal plutonium production are identified. Research reactors, mainly heavy water or graphite moderated reactors, have been used by some countries to produce weapon-grade plutonium. To prevent nuclear weapons proliferation, the Nuclear Suppliers' Group has identified nuclear reactor equipment and technology that is of importance to export control. This equipment and technology has also been implemented in the EU-regulation 428/2009. The equipment and technology that can be used in nuclear reactor applications and possible indicators on a nuclear reactor in operation is described in the report. The differences between a power reactor and a research reactor concerning these areas are high-lighted

  13. The future role of research reactors

    International Nuclear Information System (INIS)

    The decline of neutron source capacity in the next decades urges for the planning and construction of new neutron sources for basic and applied research with neutrons. Modern safety precautions of research reactors make them competitive with other ways of neutron production using non-chain reactions for many applications. Research reactors consequently optimized offer a very broad range of possible applications in basic and applied research. Research reactors at universities also in the future have to play an important role in education and training in basic and applied nuclear science. (orig.)

  14. Developments in the regulation of research reactors

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

  15. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  16. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  17. The use and evolution of the CEA research reactors

    International Nuclear Information System (INIS)

    The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of the behaviour of

  18. Baseline Graphite Characterization: First Billet

    Energy Technology Data Exchange (ETDEWEB)

    Mark C. Carroll; Joe Lords; David Rohrbaugh

    2010-09-01

    The Next Generation Nuclear Plant Project Graphite Research and Development program is currently establishing the safe operating envelope of graphite core components for a very high temperature reactor design. To meet this goal, the program is generating the extensive amount of quantitative data necessary for predicting the behavior and operating performance of the available nuclear graphite grades. In order determine the in-service behavior of the graphite for the latest proposed designs, two main programs are underway. The first, the Advanced Graphite Creep (AGC) program, is a set of experiments that are designed to evaluate the irradiated properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences, and compressive loads. Despite the aggressive experimental matrix that comprises the set of AGC test runs, a limited amount of data can be generated based upon the availability of space within the Advanced Test Reactor and the geometric constraints placed on the AGC specimens that will be inserted. In order to supplement the AGC data set, the Baseline Graphite Characterization program will endeavor to provide supplemental data that will characterize the inherent property variability in nuclear-grade graphite without the testing constraints of the AGC program. This variability in properties is a natural artifact of graphite due to the geologic raw materials that are utilized in its production. This variability will be quantified not only within a single billet of as-produced graphite, but also from billets within a single lot, billets from different lots of the same grade, and across different billets of the numerous grades of nuclear graphite that are presently available. The thorough understanding of this variability will provide added detail to the irradiated property data, and provide a more thorough understanding of the behavior of graphite that will be used in reactor design and licensing. This report covers the

  19. Software development for research reactors

    International Nuclear Information System (INIS)

    The Texas A and M University Nuclear Science Center, in a program jointly sponsored with the International Atomic Energy Agency, is developing a series of computer software programs of use at research reactor facilities. The programs cover a wide range of topics including activation and shielding calculations, control rod calibrations, power calorimetrics, and fuel inventory including burnup. Many of the programs are modified and improved versions of programs already in use at the NSC that ran on outdated computing equipment. All of the new versions were written in Fortran77 on the NSC's new TI Pro microcomputer and are IBM-compatible. This paper describes the development and translation efforts in preparing the programs for use by other facilities, and gives an overview of the aim of the development effort. A brief description of each program that has been or is to be written is given including the required inputs and the resulting outputs. This paper also addresses the original needs that brought about the development program and the benefits to facility operations that each program provides. The programs discussed are available to interested parties in a hard-copy listing as requested. (author)

  20. The concept of a research fusion reactor

    International Nuclear Information System (INIS)

    Thus,for advancement towards a commercial fusion reactor,we have proposed here as a next step a steady state operated research fusion reactor with an increased plasma-wall detachment so as to further guarantee not only the production but also a long-term (for many years) confinement of a self-sustained plasma at the existing technology level. We consider the primary goal of the research fusion reactor is the provision of full-scale conditions for carrying out materials science experiments to create and test 1 st wall materials for the commercial fusion reactor

  1. RA Research nuclear reactor - Annual report 1987

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  2. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  3. Research nuclear reactor RA - Annual Report 1989

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  4. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49

  5. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  6. Research reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 13,000 records of publications concerned with research and technology in the field of research and experimental reactors which are included in the INIS Bibliographic Database for the period from 1970 to 2001. The main objectives of this bibliometric study were: to make an inventory of research reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of research reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in research reactors research and technology. (author)

  7. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  8. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  9. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  10. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  11. IAEA Guidelines for New Research Reactor Projects

    International Nuclear Information System (INIS)

    In recent years, the interest of the IAEA Member States in developing research reactor (RR) programmes has been steadily growing. Currently a number of Member States are in different stages of new research reactor projects. Some of these Member States are building their first research reactor as their country's introduction to nuclear science and technology infrastructure. To support Member States in such efforts, the IAEA published in 2012 a Nuclear Energy Series Report NP-T-5.1: 'Specific Considerations and Milestones for a New Research Reactor Project'. This publication provides guidance on the timely preparation of a research reactor project through a sequential development process. It includes a detailed description of the range of infrastructure issues that need to be addressed and the expected level of achievement (or milestones) at the end of each phase of the project. The publication provides a discussion of the mechanisms for justification of a research reactor, and for building stakeholder support. It includes both the technical, legal, regulatory and safety infrastructure, and the development of qualified human resources needed for a research reactor. The publication also addresses the evolution of infrastructure needs from the time a Member State first considers a research reactor and its associated facilities, through the stages of planning, bid preparation, construction, start-up, and preparation for commissioning. The subsequent stages of operation, decommissioning, spent fuel and waste management issues are addressed in the publication to the degree necessary for appropriate planning prior to research reactor commissioning. The feedback from the IAEA activities, in particular from Member States establishing their first research reactor, indicated the need for further guidance on the development of the technical specifications for the bidding process of a research reactor project. In responding to these needs, a Nuclear Energy Series Report on

  12. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  13. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  14. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  15. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MWTH, with a minimum thermal neutron flux of approx, 1013 n/cm2·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation system

  16. Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

    International Nuclear Information System (INIS)

    G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs

  17. Brookhaven segment interconnect

    International Nuclear Information System (INIS)

    We have performed a high energy physics experiment using a multisegment Brookhaven FASTBUS system. The system was composed of three crate segments and two cable segments. We discuss the segment interconnect module which permits communication between the various segments

  18. Brookhaven highlights, October 1, 1989--September 30, 1990

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H. (eds.)

    1990-01-01

    This report discusses research being conducted at Brookhaven National Laboratory. Highlights from all the department are illustrated. The main topics are on accelerator development and applications. (LSP)

  19. Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-Cooled Reactor Power Plant

    International Nuclear Information System (INIS)

    In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR) graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM). A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plant lifetimes in the UK by reducing uncertainties. (author)

  20. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  1. Research reactor utilization in chemistry programmes

    International Nuclear Information System (INIS)

    The establishment and roles of the Philippines Atomic Energy Commission in promoting and regulating the use of atomic energy are explained. The research reactor, PRR-1 is being converted to TRIGA to meet the increasing demands of high-flux. The activities of PAEC in chemistry research programs utilizing reactor are discussed in detail. The current and future plans of Research and Development programs are also included. (A.J.)

  2. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  3. Operational experience with research reactors in Trombay

    International Nuclear Information System (INIS)

    The research reactors Apsara and Cirus, located at the Bhabha Atomic Research Centre, Trombay, Bombay have recently completed 30 years and 26 years of successful operation respectively. Both reactors have been used extensively for research, isotope production and manpower training. Several measures have been taken towards achieving this long successful operation. These include preventive maintenance, meticulous control on chemistry of fluid systems, proper physics management, surveillance measures and modifications to system and equipment based on experience. Extensive training of O and M personnel has been another important factor contributing towards this. Major topics detailed in this paper include detection of one leaky reactor vessel lattice tube in Cirus and its successful plugging, replacement of old control system hardware consisting of vacuum tubes with modern solid state devices in both the reactors, installation of S.S.liner in Apsara reactor pool and certain modifications towards improving fuel performance in Cirus. Some aspects of preventive maintenance are also dealt with. (author)

  4. Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Rough laminar pyrolytic carbon coating (RLPyC) is prepared by a fixed-bed method. • The salt-infiltration into IG-110 is 13.5%, less than 0.01% of RLPyC under 1.5 atm. • The helium diffusion coefficient of RLPyC coated graphite is 2.16 × 10−8 cm2/s. • The coated graphite can inhibit the liquid fluoride salt and Xe135 penetration. - Abstract: A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 105 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 105 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 105 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10−12 m2/s, much less than 1.21 × 10−6 m2/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe135 penetration

  5. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  6. Fusion reactor materials research in China

    International Nuclear Information System (INIS)

    The fusion materials research in China is introduced. Many kinds of structural materials (such as Ti-modified stainless steel, ferritic steel, HT-9, HT-7, oxide dispersion strengthening ferritic steel), tritium breeders (lithium, Li2O, γ-LiAlO2) and plasma facing materials (PFMs) (graphite with TiC and SiC coatings) have been developed or being developed. A systematic research activities on irradiation effects, compatibility, plasma materials interaction, thermal shock during disruption, tritium production, release and permeation, neutron multiplication in Be and Pb, etc. have been performed. The research activities are summarized and some experimental results are also given

  7. Safety of Ghana Research Reactor (GHARR-1)

    International Nuclear Information System (INIS)

    The Ghana Research Reactor, GHARR-1 is a low power research rector with maximum thermal power lever of 30kW. The reactor is inherently safe and uses highly enriched uranium (HEU) as fuel, light water as moderator and beryllium as a reflector. The construction, commissioning and operation of this reactor have been subjected to the system of authorization and inspection developed by the Regulatory Authority, the Radiation Protection Board (RPB) with the assistance of the International Atomic Energy Agency. The reactor has been regulated by the preparation of an Interim Safety Analysis Report (SAR) based upon International Atomic Energy Agency standards. An International Safety Assessment peer review and safe inspections have confirmed a high level of operational safety of the reactor since it started operation in 1994. Since its operation there has been no significant reported incident/accidents. Several studies have validated the inherent safety of the reactor. The reactor has been used for neutron activation analysis of various samples, research and teaching. About 1000 samples are analysed annually. The final Safety Analysis Report (SAR) was submitted (after five years of extensive research on the operational reactor) to the Regulatory Authority for review in June 2000. (author)

  8. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  9. No small fry: Decommissioning research reactors

    International Nuclear Information System (INIS)

    To get a permit to build a research reactor, would-be operators need to submit an initial decommissioning plan for the eventual shutdown of their new facility. This, however, was not a requirement back in the 1950s, 60s and 70s when most research reactors that are now nearing the end of their working lives were built. The result: many unused reactors sit idle in the middle of university campuses, research parks and hospital compounds, because their operators lack the proper plans to decommission them

  10. Brookhaven fastbus/unibus interface

    Energy Technology Data Exchange (ETDEWEB)

    Benenson, G.; Bauernfeind, J.; Larsen, R.C.; Leipuner, L.B.; Morse, W.M.; Adair, R.K.; Black, J.K.; Campbell, S.R.; Kasha, H.; Schmidt, M.P.

    1983-01-01

    A typical high energy physics experiment requires both a high speed data acquisition and processing system, for data collection and reduction; and a general purpose computer to handle further reduction, bookkeeping and mass storage. Broad differences in architecture, format or technology, will often exist between these two systems, and interface design can become a formidable task. The PDP-11 series minicomputer is widely used in physics research, and the Brookhaven FASTBUS is the only standard high speed data acquisition system which is fully implemented in a current high energy physics experiment. This paper will describe the design and operation of an interface between these two systems. The major issues are elucidated by a preliminary discussion on the basic principles of Bus Systems, and their application to Brookhaven FASTBUS and UNIBUS.

  11. Brookhaven fastbus/unibus interface

    International Nuclear Information System (INIS)

    A typical high energy physics experiment requires both a high speed data acquisition and processing system, for data collection and reduction; and a general purpose computer to handle further reduction, bookkeeping and mass storage. Broad differences in architecture, format or technology, will often exist between these two systems, and interface design can become a formidable task. The PDP-11 series minicomputer is widely used in physics research, and the Brookhaven FASTBUS is the only standard high speed data acquisition system which is fully implemented in a current high energy physics experiment. This paper will describe the design and operation of an interface between these two systems. The major issues are elucidated by a preliminary discussion on the basic principles of Bus Systems, and their application to Brookhaven FASTBUS and UNIBUS

  12. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

  13. The first university research reactor in India

    International Nuclear Information System (INIS)

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  14. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  15. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  16. Higher power density TRIGA research reactors

    International Nuclear Information System (INIS)

    The uranium zirconium hydride (U-ZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, good performance, economy of operation, and its acceptance worldwide. Of the 65 TRIGA reactors or TRIGA fueled reactors, several are located in hospitals or hospital complexes and in buildings that house university classrooms. These examples are a tribute to the high degree of safety of the operating TRIGA reactor. In the early days, the majority of the TRIGA reactors had power levels in the range from 10 to 250 kW, many with pulsing capability. An additional number had power levels up to 1 MW. By the late 1970's, seven TRIGA reactors with power levels up to 2 MW had been installed. A reduction in the rate of worldwide construction of new research reactors set in during the mid 1970's but construction of occasional research reactors has continued until the present. Performance of higher power TRIGA reactors are presented as well as the operation of higher power density reactor cores. The extremely safe TRIGA fuel, including the more recent TRIGA LEU fuel, offers a wide range of possible reactor configurations. A long core life is assured through the use of a burnable poison in the TRIGA LEU fuel. In those instances where large neutron fluxes are desired but relatively low power levels are also desired, the 19-rod hexagonal array of small diameter fuel rods offers exciting possibilities. The small diameter fuel rods have provided extremely long and trouble-free operation in the Romanian 14 MW TRIGA reactor

  17. Irradiation of thorium based fuels at research reactors of Trombay

    International Nuclear Information System (INIS)

    Utilization of large thorium reserve of our country for power production had been a prime goal from the inception of our three stage nuclear power programme. This long term policy is well reflected in irradiation programme of Thoria and Thorium rods at research reactors Cirus and Dhruva, located at Trombay. The first batch of Thorium/Thoria rods were loaded in Cirus reactor on August 28, 1960, immediately after attaining first criticality on July 10, 1960 which reflects the priority the Thorium utilization programme received from our planners. The programme was aimed for gaining sufficient experience with Thorium fuel cycle by the time our first stage of nuclear power programme attained maturity. In Cirus, these rods are irradiated in the annular gap, called J-rod annulus, between the two graphite reflectors around the pile. About 200 Thorium/Thoria rods had been irradiated in Cirus before the reactor was shut down permanently on 31st December 2010. A few Thoria assemblies have also been irradiated in fuel positions of Dhruva reactor. Apart from these, a few fuel assemblies made of ThO2, PuO2 and UO2 were irradiated in Pressurized Water Loop of Cirus to study fuel and clad behavior under high temperature and pressure, before their induction in power programme. Presently, fuel cluster with AHWR type Th-Pu and Th-LEU MOX fuel pins are being irradiated in a regular fuel position of Dhruva for obtaining vital information and experience related to Thorium based MOX fuel cycle. This paper highlights the experience gained in irradiation of Thoria/Thorium rods, expected yield of 233U, contamination level of 232U and handling of these assemblies at research reactors Cirus and Dhruva. (author)

  18. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

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

  19. Conceptual design of multipurpose compact research reactor

    International Nuclear Information System (INIS)

    Conceptual design of the high-performance and low-cost multipurpose compact research reactor which will be expected to construct in the nuclear power plant introduction countries, started from 2010 in JAEA and nuclear-related companies in Japan. The aims of this conceptual design are to achieve highly safe reactor, economical design, high availability factor and advanced irradiation utilization. One of the basic reactor concept was determined as swimming pool type, thermal power of 10MW and water cooled and moderated reactor with plate type fuel element same as the JMTR. It is expected that the research reactors are used for human resource development, progress of the science and technology, expansion of industry use, lifetime extension of LWRs and so on. (author)

  20. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  1. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 3; Trench 1 at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed Final Status Survey (FSS) of the 42-inch duct and 14-inch line in Trench 1 from Building 801 to the Stack. Sample results have been submitted as required to demonstrate that the cleanup goal of (le)15 mrem/yr above background to a resident in 50 years has been met. Four rounds of sampling, from pre-excavation to FSS, were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the U.S. Department of Energy (DOE) to perform independent verifications of decontamination and decommissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task for the HFBR Underground Utilities. ORISE, together with DOE, determined that a Type A verification of Trench 1 was appropriate based on recent verification results from Trenches 2, 3, 4, and 5, and the minimal potential for residual radioactivity in the area. The removal of underground utilities has been performed in three stages to decommission the HFBR facility and support structures. Phase 3 of this project included the removal of at least 200 feet of 36-inch to 42-inch duct from the west side to the south side of Building 801, and the 14-inch diameter Acid Waste Line that spanned from 801 to the Stack within Trench 1. Based on the pre-excavation sample results of the soil overburden, the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the gamma spectroscopy results for 14 FSS soil samples, four core samples, and one duplicate sample collected from Trench 1. Sample results for the radionuclides of concern were below the established cleanup goals. However, in sample PH-3

  2. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    These technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded

  3. Refurbishment of research reactor IRT in Sofia

    International Nuclear Information System (INIS)

    A decommissioning strategy of the research reactor IRT-Sofia is subjected to a refurbishment into critical assembly is presented. The main stages in the planning and management of a partial decommissioning are exposed. (author)

  4. Problems and experience of research reactor decommissioning

    International Nuclear Information System (INIS)

    According to the IAEA research reactor database there are about 300 research reactors worldwide. At present above 30% of them have lifetime more than 35 years, 60% - more then 25 years. After the Chernobyl accident significant efforts have been made by many countries to modernize old research reactors aiming, first of all, at ensuring of its safe operation. However, a large number of aging research reactor will be facing shutdown in the near future. Before developing the design and planning of the works it is necessary to define the concept of the reactor decommissioning. It is defined by the time of the beginning of dismantling works after the reactor shutdown and the finite state of the reactor site.The concept of the reactor decommissioning provides 3 variants in a general case: reactor conservation, or partial dismantling, or complete dismantling to 'green field' state. Specialists of three International institutions (European Commission, IAEA and the Nuclear Energy Agency/Organization for Economic Cooperation and Development) have developed a detailed plan of all actions and operations on nuclear power plants decommissioning in the framework of a joint project for cost assessment. For the reactor decontamination the following main constructions, equipment and devices are necessary: temporary storage facility for the spent fuel; general site-dismantling equipment including manipulators and 'hot' cells; facilities for 'active' equipment, personnel, tooling and washing decontamination; equipment for concentration of liquid and compactness of solid radioactive waste; temporary storage facility for radioactive waste; instrumentation and radiometric devices including , α,β,γ-spectrometers; transportable containers and other means for transportation of fuel and radioactive materials

  5. Research reactor coalitions - Second year progress report

    International Nuclear Information System (INIS)

    The IAEA, in line with its statute and mandatory responsibilities to support its member states in the promotion of peaceful uses of nuclear energy, has an initiative to promote the formation of coalitions of research reactor operators and stakeholders. These networks of research reactors are conducting joint research or other shared activities, have the potential to increase research reactor utilization and thus to improve sustainability at the same time enhancing nuclear material security and non-proliferation objectives. This effort builds upon existing IAEA efforts to enhance research reactor strategic planning, to encourage formation of research reactor networks, and to promote regional and international cooperation between research reactors. The paper will describe the Agency's progress in the second year of activities to assist in the formation of research reactor coalitions. The paper will describe the Agency's efforts in serving a catalytic and 'match-making' role for the formation of new the coalition relationships, and its activities in organizing various missions and meetings for exploratory and organizational discussions on possible coalitions and networks. The paper presents the concrete progress that has been made during the past year, including new coalitions in Eastern Europe, the Caribbean, Latin America and Central Asia. These coalitions cover a wide range of activities, for example, enhancing the regional infrastructure and capabilities for neutron sciences, developing new supplies of medicinal radioisotopes, and expanding the reach of reactor physics training courses. The paper also outlines the path forward that has been established for 2009 to support these coalitions as they mature and develop toward self-sufficiency. (author)

  6. Nuclear Research Center IRT reactor dynamics calculation

    International Nuclear Information System (INIS)

    The main features of the code DIRT, for dynamical calculations are described in the paper. With the results obtained by the program, an analysis of the dynamic behaviour of the Research Reactor IRT of the Nuclear Research Center (CIN) is performed. Different transitories were considered such as variation of the system reactivity, coolant inlet temperature variation and also variations of the coolant velocity through the reactor core. 3 refs

  7. 2012 CHEMISTRY & PHYSICS OF GRAPHITIC CARBON MATERIALS GORDON RESEARCH CONFERENCE, JUNE 17-22, 2012

    Energy Technology Data Exchange (ETDEWEB)

    Fertig, Herbert

    2012-06-22

    This conference will highlight the urgency for research on graphitic carbon materials and gather scientists in physics, chemistry, and engineering to tackle the challenges in this field. The conference will focus on scalable synthesis, characterization, novel physical and electronic properties, structure-properties relationship studies, and new applications of the carbon materials. Contributors

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

    International Nuclear Information System (INIS)

    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

  9. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

  10. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the range of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need for restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

  11. Monte Carlo simulation of the Greek research reactor neutron irradiation positions using MCNP

    International Nuclear Information System (INIS)

    Prediction of neutron flux at the irradiation devices of a research reactor facility is essential for the design and evaluation of experiments involving material irradiations. A computational model of the Greek Research Reactor (GRR-1) was developed using the Monte Carlo code MCNP with continuous energy neutron cross-section data evaluations from ENDF/B-VI library. The model included detailed geometrical representation of the fuel and control assemblies, beryllium reflectors, irradiation devices and the graphite pile. The MCNP model was applied to predict neutron flux at the in-pool irradiation positions and the graphite pile. The MCNP estimated neutron fluxes were compared with measurements using activation foils and a good agreement between calculated and experimental results was observed. (author)

  12. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  13. Research nuclear reactor RA - Annual Report 1997

    International Nuclear Information System (INIS)

    RA reactor is not in operation since 1984, activities related to revitalisation of the RA reactor started in 1986. The planned actions related to renewal of the reactor components were finished except for the most important action, related to exchange of complete reactor instrumentation which was delayed. Only 80% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the sanctions imposed to our country. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 42 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved. Research reactor RA Annual report for year 1997 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  14. Research nuclear reactor RA - Annual Report 1996

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986, were continued in 1996. All the planned actions related to renewal of the reactor components were finished. The last, and at the same time most important action, related to exchange of complete reactor instrumentation is underway, delayed. The delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 80% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 43 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved. Research reactor RA Annual report for year 1996 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  15. Status of Research Reactor Utilization in Brazil

    International Nuclear Information System (INIS)

    Brazil has four research reactors in operation: the IEA-R1, a pool type research reactor of 5 MW; the IPR-R1, a TRIGA Mark I type research reactor of 100 kW; the ARGONAUTA, an Argonaut type research reactor of 500 W; and the IPEN/MB-01 a critical facility of 100 W. Research reactor utilization has more than fifty years in Brazil. The first three reactors, constructed in the late 50's and early 60's at university campus in Sao Paulo, Belo Horizonte and Rio de Janeiro, had their utilization for training, teaching and nuclear research. The IPEN/MB-01, designed and constructed in IPEN in the late 80's, is utilized for the development and qualification of reactor physics calculation for PWR core application. The IEA-R1 has had its application and utilization increased through the years and it is presently used for radioisotope production, neutron beam application, neutrongraphy, neutron activation analysis, and limited fuel and material irradiation tests, besides the regular use for training and teaching. The low power of the reactor and the lack of hot cells for post irradiation analysis limits its technical application for the nuclear fuel industry. Brazil has two nuclear power plants in operation, one unit starting construction and four more units planned for the next two decades. Brazil has significant quantities of uranium ore and has expertise in all the fuel cycle steps, including uranium enrichment, and produces the fuel assemblies for the nuclear power plants. These industrial activities demand the need of material and fuel irradiation tests. IPEN produces radiopharmaceutical kits for the treatment of more than three million patients each year. The majority of the radiopharmaceutical kits is produced from imported radioisotopes. The increasing price and shortage of world supply of 99mTc leads also to the need of increasing the radioisotope production in Brazil. Due to these new demands, the Brazilian Nuclear Energy Commission is analyzing the costs and benefits

  16. Utilization of the Research Reactor ASTRA

    International Nuclear Information System (INIS)

    A short history and an overview over present research activities at the 10 MW Pool Type Reactor ASTRA of the Austrian Research Centre Seibersdorf are given. The projects comprise: medical and industrial isotope production, material irradiations (e.g. silicon doping), neutron activation analysis, geological dating and radiation induced mutation techniques for agricultural research. (author)

  17. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    Energy Technology Data Exchange (ETDEWEB)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-06-10

    The aim of the research presented here was to identify the chemical form of 14C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14C, with a half-life of 5730 years.

  18. Research reactors in Kazakhstan: Conditions, safety and utilization

    International Nuclear Information System (INIS)

    Full text: There are 4 research reactors and 1 critical assembly in Kazakhstan. Three of them: RA, IGR and IWG.1M reactors belong to the Institute of Atomic Energy. The last WWR-K reactor and the critical assembly belong to the Institute of Nuclear Physics. Both institutes are parts of the National Nuclear Centre of the Republic of Kazakhstan. And all of these reactors are under the supervising of the regulatory body the Atomic Energy Committee of the Republic of Kazakhstan (KAEC). One of them is RA reactor and placed near Kurchatov town on the 'Baikal-1' site at former Semipalatinsk Nuclear Testing Site. Nowadays it is in extended shutdown regime and defueling. IGR reactor is impulse type and has clumped Uranium-graphite 90% enrichment fuel. It started to operate in 1961 and placed near Kurchatov town on the site at former Semipalatinsk Nuclear Testing Site. The next IWG.1M reactor has 35 MWth power and is working in long impulse regime. It placed near Kurchatov town on the 'Baikal-1' site at former Semipalatinsk Nuclear Testing Site too. It has 90% enrichment Uranium-Zirconium fuel and started in operation in 1975. The last one WWR-K reactor has 6 MWth power and is working in stationary regime during 20 days campaign. It is placed near Almaty city with the about 1.5 billion population. It has 36% enrichment Uranium-Aluminum fuel and started in operation in 1967. The critical assembly is the prototype of the WWR-K reactor and has the same fuel type as the reactor has. It is placed at the same reactor building and started in operation in 1972 and modified in 1982. It is obviously see that Kazakhstan has 4 different type reactors with high enrichment different type fuel and all of them have the age about 30 years approaching final shutdown. So the main problem to the regulatory body KAEC is to estimate their conditions and the limits of their safety operation. Every three years the special commissions review the conditions of each reactor and make a decision to its

  19. Material test reactor fuel research at the BR2 reactor

    International Nuclear Information System (INIS)

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

  20. Experience in Neutron Physics Acquired at Marcoule and Chinon. Its Value for the Graphite-Reactor Programme

    International Nuclear Information System (INIS)

    The entry into service of the first French power reactor - Gl, G2 and G3 at Marcoule and EDF1 at Chinon - has provided fundamental experience for the further development of this reactor type. This experience has accrued both from start-up tests and from power operation. The most important start-up tests consisted of : (a) Progressive replacement experiments, which made it possible to perfect the methods of calculation for G2 and EDF1; and (b) Fixed absorber and control rod-tests. Through the operation of G2 and G3, a better adjustment of the reactivity balance under power has been achieved, taking into account effects of temperature and xenon poisoning. Similarly, experiments carried out with the reactors during operation under power have made it possible to check the validity of the kinetic models used to study transient phenomena in this type of reactor. The experience acquired with the first French graphite reactors - though it still has to be completed by measurements of a more basic and systematic kind on assemblies specially designed for the purpose, such as MARIUS and CESAR - has thus proved to be of extreme value both for future projects and for studies on the operation of large power stations, where problems of control and kinetics are especially important. (author)

  1. Degradations of thermal shock resistance and fracture toughness of neutron irradiated reactor graphite

    International Nuclear Information System (INIS)

    This paper deals with the irradiation effects on the thermal shock resistance Δ = σtk/Eα (σt: tensile strength, k: thermal conductivity, E: Young's modulus, α: thermal expansivity) and the thermal shock fracture toughness ∇ = KIck/Eα (KIc: fracture toughness value of mode I) in addition to other mechanical properties such as the diametral compressive strength and fracture toughness of two kinds of near isotropic graphite neutron irradiated at 750 ∼ 1,000 deg C to a dose of (1.1 ∼ 1.5) x 1021 n/cm2 ( > 29 fJ). One of the graphite specimen is an isostatically molded graphite IG-11 with fine grain petroleum coke and the other is a binderless molded graphite HCB-18 which was prepared using mesophase pitch carbon with very fine grain size. These measurements are carried out by means of disk testing method developed by us. Results show that both the thermal shock resistance and the thermal shock fracture toughness of the two kinds of graphite after irradiation decrease considerably in contrast with increasing trends of the usual mechanical strengths. (author)

  2. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

  3. Thermal hydraulic analysis of nuclear research reactors

    International Nuclear Information System (INIS)

    A loss of coolant accident (LOCA) can cause total or partial core uncovery which is followed by substantial fuel element temperature increase due to fuel residual heat. It is essential to demonstrate that such a temperature increase does not lead to excessive core melting and to significant radioactive material release into the reactor building and consequently to the environment. The THEAP computer codes able to perform reliable analysis of such accidents have been developed. THEAP-I is a computer code developed with the aim to contribute to the safety analysis of the MTR open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. The mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is included, referring to the Greek Research Reactor analysis, under a hypothetical severe loss of coolant accident. The micro computer version of the code is also described. More emphasis is given in the new features of the code (i.e. input data structure). A set of instructions for running in an IBM-AT2 computer with the microsoft FORTRAN V4.0 is included together with a sample problem referring to the Greek Research Reactor. THEAP-I can be used also for other MTR open pool research reactors. Refs and figs

  4. Description of the Korean multipurpose research reactor

    International Nuclear Information System (INIS)

    The Korean Multipurpose Research Reactor (KMRR) was still under construction at the time of the conference, and was scheduled for completion in 1994. It has since been completed. Owned by Korean Atomic Energy Research Institute (KAERI), KMRR is designed for fuel and materials testing, production of key nuclides (including 99Mo, 131I, 192Ir, 60Co, and transmutation doped Si), neutron activation analysis, and neutron radiography. KMRR is a 30 MWth open pool type research reactor, with a forced upward light water moderator and coolant flow, and a heavy water annular reflector. The fuel is made from low-enriched U-Si-Al alloy, with finned aluminum cladding. Two configurations of fuel bundle have been designed, namely, hexagonal 36 element and circular 18 element. The reactor has seven tangential beam tubes, the position of which has been carefully selected to maximize thermal neutron flux while minimizing fast neutrons and gamma. Heat is transported to two heat exchangers by the primary coolant circuit, and thence by the secondary coolant circuit to a set of cooling towers. The reactor regulating system has two stepping motors that drive four hafnium control rods. The control functions are performed by two redundant programmable controllers. The reactor protection system is equipped with four hafnium shutoff rods driven by a fail-safe hydraulic circuit. The design of KMRR uses leading edge technology, and it might well figure among the best multipurpose research reactors in the world. 5 figs., 1 tab

  5. Evaluation of graphite/steam interactions for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    In this report we present the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during a coolant inleakage accident in ITER. This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We have measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 C and 1700 C. The effects of steam flow rate, and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We have used the relationships for GraphNOL N3M in a thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 20 refs., 14 figs., 1 tab

  6. Investigation of the cyclone model for helium cleaning system from graphite dust at helium-cooled reactors

    International Nuclear Information System (INIS)

    For maintaining in a gas circuit of the high-temperature pebble-bed reactor of minimum graphite dust concentration, the two-stage cleanup system is foreseen: Cyclones and filters of fine cleaning. The results of cyclone investigation at the experimental plant representing a closed circuit with heat exchange equipment, isolation valves, measuring devices and simulated section with 0.4 MPa working pressure, temperature about 300 deg C. The experiments have been conducted under 0.1-0.4 MPa pressure. In case of helium humidity in the circuit about 0.1%, dust conglomerate production is not observed. The investigation has been performed at different helium temperatures: 20 deg C, 170 deg C, 300 deg C. At all stages the investigations have been performed at 100-180 m3h - 0.1-0.4 MPa pressure, and. fraction compositions of graphite dust from 0 to 50 μm and from 0 to 90 μm. The computer statistical processing of the experiment results has been performed by tthe EG-1030 computer. The performed investigations allow one to recommend using cyclones as the first stage of the helium coolant cleaning in the plants with gas cooled reactors

  7. Status report of Indonesian research reactor

    International Nuclear Information System (INIS)

    A general description of three Indonesian research reactor, its irradiation facilities and its future prospect are described. Since 1965 Triga Mark II 250 KW Bandung, has been in operation and in 1972 the design powers were increased to 1000 KW. Using core grid form Triga 250 KW BATAN has designed and constructed Kartini Reactor in Yogyakarta which started its operation in 1979. Both of this Triga type reactors have served a wide spectrum of utilization such as training manpower in nuclear engineering, radiochemistry, isotope production and beam research in solid state physics. Each of this reactor have strong cooperation with Bandung Institute of Technology at Bandung and Gajah Mada University at Yogyakarta which has a faculty of Nuclear Engineering. Since 1976 the idea to have high flux reactor has been foreseen appropriate to Indonesian intention to prepare infrastructure for nuclear industry for both energy and non-energy related activities. The idea come to realization with the first criticality of RSG-GAS (Multipurpose Reactor G.A. Siwabessy) in July 1987 at PUSPIPTEK Serpong area. It is expected that by early 1992 the reactor will reached its full power of 30 MW and by end 1992 its expected that irradiation facilities will be utilized in the future for nuclear scientific and engineering work. (author)

  8. THE NEXT GENERATION NUCLEAR PLANT GRAPHITE PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    William E. Windes; Timothy D. Burchell; Robert L. Bratton

    2008-09-01

    Developing new nuclear grades of graphite used in the core of a High Temperature Gas-cooled Reactor (HTGR) is one of the critical development activities being pursued within the Next Generation Nuclear Plant (NGNP) program. Graphite’s thermal stability (in an inert gas environment), high compressive strength, fabricability, and cost effective price make it an ideal core structural material for the HTGR reactor design. While the general characteristics necessary for producing nuclear grade graphite are understood, historical “nuclear” grades no longer exist. New grades must be fabricated, characterized, and irradiated to demonstrate that current grades of graphite exhibit acceptable non-irradiated and irradiated properties upon which the thermo-mechanical design of the structural graphite in NGNP is based. The NGNP graphite R&D program has selected a handful of commercially available types for research and development activities necessary to qualify this nuclear grade graphite for use within the NGNP reactor. These activities fall within five primary areas; 1) material property characterization, 2) irradiated material property characterization, 3) modeling, and 4) ASTM test development, and 5) ASME code development efforts. Individual research and development activities within each area are being pursued with the ultimate goal of obtaining a commercial operating license for the nuclear graphite from the US NRC.

  9. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

    The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

  10. BNCT activities at Slovenian TRIGA research reactor

    International Nuclear Information System (INIS)

    It has been reported that satisfactory thermal/epithermal neutron beams for Boron Neutron Capture Therapy (BNCT) could be designed at TRIGA research reactors These reactors are generally perceived as being safe to install and operate in populated areas. This contribution presents the most recent BNCT research activities on the 'Jozef Stefan' Institute, where epithermal neutron beam for 'in-vitro' irradiation has been developed and experimentally verified. Furthermore, The Monte Carlo feasibility study of development of the epithermal neutron beam for BNCT clinical trials of human patients in thermalising column (TC) of TRIGA reactor has been carried out. The simulation results prove, that a BNCT irradiation facility with performances, comparable to existing beam throughout the world, could be installed in TC of the TRIGA reactor. (author)

  11. Nuclear reactor safety research in Kazakhstan

    International Nuclear Information System (INIS)

    Full text : The paper summarizes activities being implemented by the National Nuclear Center of the Republic of Kazakhstan in support of safe operation of nuclear reactors; shows its crucial efforts and further road map in this line. As is known, the world community considers nuclear reactor safety as one of the urgent research areas. Kazakhstan has been pursuing studies in support of nuclear energy safety since early 80s. The findings allow to coordinate available computational methods and design new ones while validating new NPP Projects and making analysis for reactor installations available

  12. Compressive impact strength of high temperature gas-cooled reactor graphite

    International Nuclear Information System (INIS)

    To investigate the effect of strain rate on fracture behavior for coarse grained nuclear graphite, PGX, a hydraulic servo type impact testing machine has been constructed and compressive impact strength test was performed at various strain up to more than 100(1/s). From the results, the following conclusions were derived. (1) Compressive impact strength of graphite increases with increasing of strain rate in the range of 10-3 to 100(1/s). (2) Compressive impact strength decreases drastically for strain rates more than 100(1/s). (3) Compressive impact strength dose not depend on specimen volume. (author)

  13. Making better use of research reactors

    International Nuclear Information System (INIS)

    Some 250 research reactors are in operation in the world today, and there are problems in putting them to the most fruitful use. The difficulties - of trained manpower, of auxiliary equipment, of satisfactory research programmes, of co-ordination, between the various disciplines - are common to all users. But as is only to be expected, they press more heavily on the newly-established centres, particularly those in the developing countries which are lacking in long experience in research and usually severely limited as to technical manpower and money. The IAEA has been turning its attention to this question for the past three or four years - ever since, in fact, its early assistance missions and other field operations brought it into close contact with the operations of numerous Member States. The task of providing assistance and advice in this matter is growing. Many centres have been building research reactors under bilateral arrangements; with the completion of their projects this form of aid usually ends, and they look to IAEA for help in operating the reactors. Although some critics consider that difficulties have been caused by premature construction of research reactors, before well-founded programmes of nuclear research had been developed in the countries concerned, several valid motives have led to the establishment of some of these centres at an early stage. A research reactor often provides an effective stimulant for scientific research in the country. It is a remarkably versatile tool for workers in many fields of science and technology. There have been instances where the establishment of a research reactor has had a great impact on the scientific education of a country and has led to a salutary reappraisal and reforms. A reactor is sometimes considered to be a particularly effective means of retaining in the country men trained in the nuclear field. This particular problem is common to most countries. In fact, it is a feature of the present age that

  14. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    For a period of about 10 years AECL had a significant program looking into the possibility of developing U3Si as a high density replacement for the UO2 pellet fuel in use in CANDU power reactors. The element design consisted of a Zircaloy-clad U3Si rod containing suitable voidage to accommodate swelling. We found that the binary U3Si could not meet the defect criterion for our power reactors, i.e., one month in 300 degree C water with a defect in the sheath and no significant damage to the element. Since U3Si could not do the job, a new corrosion resistant ternary U-Si-Al alloy was developed and patented. Fuel elements containing this alloy came close to meeting the defect criterion and showed slightly better irradiation stability than U3Si. Shortly after this, the program was terminated for other reasons. We have made much of this experience available to the Low Enrichment Fuel Development Program and will be glad to supply further data to assist this program

  15. The distribution of the reactor's graphite resource in cells of the operation and protection system with the degradation of the thermo-physical properties taking into account

    International Nuclear Information System (INIS)

    The evaluation of the graphite service life of the tube water-cooling reactors with the graphite moderator is carried out. The changes of the heat-exchange and heat-transfer conditions that are caused by thermophysical properties degradation and by the form changes of the structural elements are taken into account. Peculiarities that are stipulated by the spectrum of the neutron flows differences and by the differences of parameters of the gamma-radiation in the fuel channels and in the reactor control system channels are taken into account

  16. RMB. The new Brazilian multipurpose research reactor

    International Nuclear Information System (INIS)

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also presents the

  17. RMB. The new Brazilian multipurpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto; Soares, Adalberto Jose [Comissao Nacional de Energia Nuclear (CNEN) (Brazil)

    2015-01-15

    Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also

  18. Determination of research reactor fuel burnup

    International Nuclear Information System (INIS)

    This report was prepared by a Consultants Group which met during 12-15 June 1989 at the Jozef Stefan Institute, Yugoslavia, and during 11-13 July 1990 at the IAEA Headquarters in Vienna, Austria, with subsequent contributions from the Consultants. The report is intended to provide information to research reactor operators and managers on the different, most commonly used methods of determining research reactor fuel burnup: 1) reactor physics calculations, 2) measurement of reactivity effects, and 3) gamma ray spectrometry. The advantages and disadvantages of each method are discussed. References are provided to assist the reactor operator planning to establish a programme for burnup determination of fuel. Destructive techniques are not included since such techniques are expensive, time consuming, and not normally performed by the reactor operators. In this report, TRIGA fuel elements are used in most examples to describe the methods. The same techniques however can be used for research reactors which use different types of fuel elements. 22 refs, 13 figs, 2 tabs

  19. Research reactor operation, maintenance and utilization in Thailand

    International Nuclear Information System (INIS)

    Thai Research Reactor, TRR-1/Ml is currently operated at one meggawatt steady state power. It was first commissioned in 1962 as MTR swimming pool type research reactor, and the core was later changed to TRIGA Mark III. The new core has been operated since 1978. Current core configuration Core no.5, has totally 108 low-enriched-uranium (LEU) fuel elements of 8.5 wt% and 20 wt% a mixed core. It is light water cooled and moderated with end graphite and radial water reflected. Fuel meat consists of homogeneous mixture of U-ZrH1.6 alloy which provides built-in negative temperature coefficient. Since 1978, The TRIGA core has generated energy cumulatively of 358.18 MWD thermal, as of 30 September 1986. There are six major groups of facilities; four for neutron activation analysis (NAA), one for isotope production, and a group of beam tubes. The average utilization time of these facilities varies from 0 per cent to 98 per cent. Two main utilization are NAA and radioisotope production. (author)

  20. Status report of Indonesian research reactors

    International Nuclear Information System (INIS)

    A general description of the three Indonesia research reactors, their irradiation facilities and future prospect are given. The 250 kW Triga Mark II in Bandung has been in operation since 1965 and in 1972 its designed power was increased to 1000 kW. The core grid from the previous 250 kW Triga Mark II was then used by Batan for designing and constructing the Kartini reactor in Yogyakarta. This reactor commenced its operation in 1979. Both Triga reactors have served a wide spectrum of utilization such as for manpower training in nuclear engineering, radiochemistry, isotope production, and beam research in solid state physics. The Triga reactor management in Bandung has a strong cooperation with the Bandung Institute of Technology and the one in Yogyakarta with the Gadjah Mada University which has a Nuclear Engineering Department at its Faculty of Engineering. In 1976 there emerged an idea to have a high flux reactor appropriate for Indonesia's intention to prepare an infrastructure for both nuclear energy and non-energy industry era. Such an idea was then realized with the achievement of the first criticality of the RSG-GAS reactor at the Serpong area. It is now expected that by early 1992 the reactor will reach its full 30 MW power level and by the end of 1992 the irradiation facilities be utilizable fully for future scientific and engineering work. As a part of the national LEU fuel development program a study has been underway since early 1989 to convert the RSG-GAS reactor core from using oxide fuel to using higher loading silicide fuel. (author)

  1. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  2. RRFM (European Research Reactor Conference) 2011 Transactions

    International Nuclear Information System (INIS)

    The RRFM conference is an international forum for researchers, operators and decision-makers to discuss all significant aspects of Research Reactor utilisation. In order to improve operational efficiency and fuel safety and contribute to the search for back-end solutions for spent fuel

  3. Event management in research reactors

    International Nuclear Information System (INIS)

    In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

  4. Refurbishing - A cost effective option for long term operation of research reactors

    International Nuclear Information System (INIS)

    In recent times it has been observed that the life time of research reactors can be significantly extended through implementation of appropriate refurbishing actions. This is feasible since the initially proclaimed design life of such facilities in most cases, is an arbitrary number. This aspect is applicable to research reactors, nuclear power plants and many nuclear fuel cycle facilities. In India extensive refurbishing of the 40 MWt research reactor CIRUS was done during 1997-2002. Cirus is a vertical tank type reactor using natural uranium as fuel, light water as coolant, heavy water as moderator and graphite as reflector and became operational in 1960. Detailed ageing assessment of its systems, structures and components was done during 1992-1997 after the reactor had been in operation for about 30 years. The assessment included inspections, operating experience review, review of the Safety Analysis Report, seismic re-evaluation of structures and assessment of stored energy in the graphite reflector. A refurbishing plan was then drawn-up and preparatory work undertaken that included development of procedures, procurement of replacement items etc. The reactor was then shutdown, core unloaded and reactor systems were prepared for preservation during refurbishing. Extensive refurbishing was then carried out and the reactor brought back into operation successfully. During refurbishing, a low temperature vacuum evaporation based desalination unit was also coupled to the reactor to serve as demonstration of using waste heat from a research reactor for sea water desalination. The scope of refurbishing of Cirus got considerably expanded as a result of identification of additional jobs during further inspections undertaken after reactor shut-down and core unloading. Consequently the refurbishing took about 5 years against the initially planned period of about 3 years. In spite of this, the facility could be refurbished at a cost which is less than 10% of the cost of

  5. Degradations of thermal shock resistance and the fracture toughness of reactor graphite subjected to neutron irradiation

    International Nuclear Information System (INIS)

    This paper deals with the irradiation effects on the thermal shock resistance Δ = σt k/Ea (σt: tensile strength, k: thermal conductivity, E: Young's modulus, a: thermal expansivity) and the thermal shock fracture toughness ∇ = KIc k/Ea (KIc: fracture toughness value of mode 1) in addition to other mechanical properties such as the diametral compressive strength and fracture toughness. Two kinds of near isotropic graphite subjected to neutron irradiations of (1.1 - 1.5) x 1021 n/cm2 (> 29 fJ) at 750 - 1000 deg C are prepared for the measurements. One is an isostatically molded graphite IG-11 with fine grain petroleum coke. The other is a binderless molded graphite HCB-18 which was prepared using mesophase pitch carbon of very fine grain size. These measurements are carried out by means of our developed disk testing method. Results show that both the thermal shock resistance and the thermal shock fracture toughness of the two graphites after irradiation are found to decrease considerably in contrast with increasing trends of the usual mechanical strengths. (author)

  6. Thermal reactor safety CNEN research programs

    International Nuclear Information System (INIS)

    A review of CNEN (National Committee for Nuclear Energy, Italy) programs in the field of thermal reactor safety research is given. The ASCOT program (research program on safety aspects of thermal reactor cores) is briefly described. ASCOT is a program aiming at studying fuel behavior under accident conditions; it is mainly focused on development and experimental testing of analytical models and computer codes relevant to thermohydraulic and mechanical behavior of fuel under transient conditions. The program, fully financed by CNEN, is carried out in CNEN laboratories, in CISE laboratories (particularly for thermohydraulic experiments) and in JRC Ispra Centre (in pile experiments, by ESSOR reactor). Other CNEN research programs in the field of water reactor safety are also described; they concern thermohydraulics and mechanics problems (model development and experimental tests on pressure suppression, ECCS, etc.) and are performed both in CNEN laboratories and in other Italian organizations, under CNEN sponsorship. A short description of some facilities used for ASCOT and other CNEN programs is given: SARA loop (a loop of ESSOR reactor, basically conceived for safety tests, including operation with failed fuel rods); CIRCE and IETI loops (CISE, large-scale facilities for thermohydraulic experiments on blow-down, ECCS, etc.); ADI (a CNEN, large-scale loop where pressure suppression experiments are performed), and so on. Finally, the report describes interesting safety researches on various types of reactors: researches on external events (seismology, etc.), radioactive effluent control (e.g., filtration, effects to environment); these researches also are carried out directly in CNEN laboratories or in other Italian organizations, under CNEN sponsorship. Information is given on a national seismological network and on other installations for these experimental researches

  7. Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Willaim Windes; G. Strydom; J. Kane; R. Smith

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  8. BROOKHAVEN: High energy gold

    International Nuclear Information System (INIS)

    On April 24, Brookhaven's Alternating Gradient Synchrotron (AGS) started to deliver gold ions at 11.4 GeV per nucleon (2,000 GeV per ion) to experimenters who were delighted not only to receive the world's highest energy gold beam but also to receive it on schedule

  9. Advanced fuel in the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovsky, I. [KFKI Atomic Energy Research Inst., Budapest (Hungary)

    1997-07-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  10. EURATOM research framework programmes on reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Deffrennes, M.; Hugon, M.; Manolatos, P.; Van Goethem, G.; Webster, S. [European Commission, DG Research J2 Nuclear Fission and Radiation Protection CDMA 1/55, Brussels (Belgium)

    2007-07-01

    The activities of the European Commission (EC) in the field of nuclear energy are governed by the Treaty establishing the European Atomic Energy Community (EURATOM). The research activities of the European Union (EU) are designed as multi-annual Framework Programmes (FP) managed by the European Commission. The EURATOM Research and Training Programmes cover both nuclear Fusion and Fission. EURATOM-FP6 supports the following projects: -) NULIFE project: Nuclear Plant Life Prediction; -) COVERS project: VVER Safety Research; -) PERFECT project: Prediction of Irradiation Damage Effects on Reactor Components; -) NURESIM project: European Platform for Nuclear Reactor Simulations; -)EC-SARNET project: Sustainable Integration of European Research on Severe Accident Phenomenology; -) RAPHAEL project: Reactor for Process Heat, Hydrogen and Electricity Generation; -)GCFR project: Gas-Cooled Fast Reactor; -) EUROTRANS project: Transmutation of HLW in ADS; -) JHR-CA project: Jules Horowitz Reactor Co-ordination Action; and NEPTUNO project: Nuclear European Platform of Training and University Organisations. Other parts of the EURATOM FP, covering Waste Management and Radiation Protection, as well as Fusion Energy, are not detailed in this paper.

  11. EURATOM research framework programmes on reactor systems

    International Nuclear Information System (INIS)

    The activities of the European Commission (EC) in the field of nuclear energy are governed by the Treaty establishing the European Atomic Energy Community (EURATOM). The research activities of the European Union (EU) are designed as multi-annual Framework Programmes (FP) managed by the European Commission. The EURATOM Research and Training Programmes cover both nuclear Fusion and Fission. EURATOM-FP6 supports the following projects: -) NULIFE project: Nuclear Plant Life Prediction; -) COVERS project: VVER Safety Research; -) PERFECT project: Prediction of Irradiation Damage Effects on Reactor Components; -) NURESIM project: European Platform for Nuclear Reactor Simulations; -)EC-SARNET project: Sustainable Integration of European Research on Severe Accident Phenomenology; -) RAPHAEL project: Reactor for Process Heat, Hydrogen and Electricity Generation; -)GCFR project: Gas-Cooled Fast Reactor; -) EUROTRANS project: Transmutation of HLW in ADS; -) JHR-CA project: Jules Horowitz Reactor Co-ordination Action; and NEPTUNO project: Nuclear European Platform of Training and University Organisations. Other parts of the EURATOM FP, covering Waste Management and Radiation Protection, as well as Fusion Energy, are not detailed in this paper

  12. Performance of a multipurpose research electrochemical reactor

    International Nuclear Information System (INIS)

    Highlights: → For this reactor configuration the current distribution is uniform. → For this reactor configuration with bipolar connection the leakage current is small. → The mass-transfer conditions are closely uniform along the electrode. → The fluidodynamic behaviour can be represented by the dispersion model. → This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of ±10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  13. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  14. Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2006-01-01

    Full Text Available RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

  15. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    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)

  16. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  17. Research nuclear reactor RA - Annual report 1992

    International Nuclear Information System (INIS)

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data

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

    International Nuclear Information System (INIS)

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

  19. Research nuclear reactor RA, Annual Report 2001

    International Nuclear Information System (INIS)

    During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001

  20. Research nuclear reactor RA - Annual Report 1994

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Activities related to improvement of Russian project were continued in 1994. Control and maintenance of the reactor components was done regularly and efficiently. Extensive repair of the secondary coolant loop is almost finished and will be completed in the first part of 1995 according to existing legal procedures and IAEA recommendations. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 47 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  1. Experience at SAPHIR Research Reactor, Switzerland

    International Nuclear Information System (INIS)

    The former SAPHIR research reactor has been dismantled completely without any significant difficulty. There are several factors underpinning the successful dismantling of SAPHIR: – Good housekeeping during operation and after shutdown; – Good maintenance of the infrastructure before and after shutdown; – Experienced personnel with knowledge of the reactor history; – Stable legal framework; – Close cooperation with the regulatory authority; – Excellent infrastructure of a large research centre; – Stable financing; – Stable organization, motivated personnel; – Support from skilful local companies; – Waste conditioning and treatment routes on-site and approved by the regulatory authority

  2. LETTER REPORT - INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) personnel visited the Brookhaven National Laboratory (BNL) on August 17 through August 23, 2010 to perform visual inspections and conduct independent measurement and sampling of the 'Outside Areas' at the High Flux Beam Reactor (HFBR) decommissioning project. During this visit, ORISE was also able to evaluate Fan House, Building 704 survey units (SUs) 4 and 5, which are part of the Underground Utilities portion of the HFBR decommissioning project. ORISE performed limited alpha plus beta scans of the remaining Fan House foundation lower walls and remaining pedestals while collecting static measurements. Scans were performed using gas proportional detectors coupled to ratemeter-scalers with audible output and encompassed an area of approximately 1 square meter around the static measurement location. Alpha plus beta scans ranged from 120 to 460 cpm. Twenty smears for gross alpha and beta activity and tritium were collected at judgmentally selected locations on the walls and pedestals of the Fan House foundation. Attention was given to joints, cracks, and penetrations when determining each sample location. Removable concentrations ranged from -0.43 to 1.73 dpm/100 cm2 for alpha and -3.64 to 7.80 dpm/100 cm2 for beta. Tritium results for smears ranged from -1.9 to 9.0 pCi/g. On the concrete pad, 100% of accessible area was scanned using a large area alpha plus beta gas proportional detector coupled to a ratemeter-scaler. Gross scan count rates ranged from 800 to 1500 cpm using the large area detector. Three concrete samples were collected from the pad primarily for tritium analysis. Tritium concentrations in concrete samples ranged from 53.3 to 127.5 pCi/g. Gamma spectroscopy results of radionuclide concentrations in concrete samples ranged from 0.02 to 0.11 pCi/g for Cs-137 and 0.19 to 0.22 pCi/g for Ra-226. High density scans for gamma radiation levels were performed in accessible areas in each SU, Fan House

  3. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  4. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  5. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

  6. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given

  7. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    International Nuclear Information System (INIS)

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U3Si2) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x1014 cm-2s-1. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  8. Brookhaven National Laboratory source water assessment for drinking water supply wells

    International Nuclear Information System (INIS)

    The BNL water supply system meets all water quality standards and has sufficient pumping and storage capacity to meet current and anticipated future operational demands. Because BNL's water supply is drawn from the shallow Upper Glacial aquifer, BNL's source water is susceptible to contamination. The quality of the water supply is being protected through (1) a comprehensive program of engineered and operational controls of existing aquifer contamination and potential sources of new contamination, (2) groundwater monitoring, and (3) potable water treatment. The BNL Source Water Assessment found that the source water for BNL's Western Well Field (comprised of Supply Wells 4, 6, and 7) has relatively few threats of contamination and identified potential sources are already being carefully managed. The source water for BNL's Eastern Well Field (comprised of Supply Wells 10, 11, and 12) has a moderate number of threats to water quality, primarily from several existing volatile organic compound and tritium plumes. The g-2 Tritium Plume and portions of the Operable Unit III VOC plume fall within the delineated source water area for the Eastern Well Field. In addition, portions of the much slower migrating strontium-90 plumes associated with the Brookhaven Graphite Research Reactor, Waste Concentration Facility and Building 650 lie within the Eastern source water area. However, the rate of travel in the aquifer for strontium-90 is about one-twentieth of that for tritium and volatile organic compounds. The Laboratory has been carefully monitoring plume migration, and has made adjustments to water supply operations. Although a number of BNL's water supply wells were impacted by VOC contamination in the late 1980s, recent routine analysis of water samples from BNL's supply wells indicate that no drinking water standards have been reached or exceeded. The high quality of the water supply strongly indicates that the operational and engineered controls implemented over the past

  9. Proceedings of RIKEN BNL Research Center Workshop: Brookhaven Summer Program on Quarkonium Production in Elementary and Heavy Ion Collisions

    Energy Technology Data Exchange (ETDEWEB)

    Dumitru, A.; Lourenco, C.; Petreczky, P.; Qiu, J., Ruan, L.

    2011-08-03

    Understanding the structure of the hadron is of fundamental importance in subatomic physics. Production of heavy quarkonia is arguably one of the most fascinating subjects in strong interaction physics. It offers unique perspectives into the formation of QCD bound states. Heavy quarkonia are among the most studied particles both theoretically and experimentally. They have been, and continue to be, the focus of measurements in all high energy colliders around the world. Because of their distinct multiple mass scales, heavy quarkonia were suggested as a probe of the hot quark-gluon matter produced in heavy-ion collisions; and their production has been one of the main subjects of the experimental heavy-ion programs at the SPS and RHIC. However, since the discovery of J/psi at Brookhaven National Laboratory and SLAC National Accelerator Laboratory over 36 years ago, theorists still have not been able to fully understand the production mechanism of heavy quarkonia, although major progresses have been made in recent years. With this in mind, a two-week program on quarkonium production was organized at BNL on June 6-17, 2011. Many new experimental data from LHC and from RHIC were presented during the program, including results from the LHC heavy ion run. To analyze and correctly interpret these measurements, and in order to quantify properties of the hot matter produced in heavy-ion collisions, it is necessary to improve our theoretical understanding of quarkonium production. Therefore, a wide range of theoretical aspects on the production mechanism in the vacuum as well as in cold nuclear and hot quark-gluon medium were discussed during the program from the controlled calculations in QCD and its effective theories such as NRQCD to various models, and to the first principle lattice calculation. The scientific program was divided into three major scientific parts: basic production mechanism for heavy quarkonium in vacuum or in high energy elementary collisions; the

  10. Fuel behavior comparison for a research reactor

    Science.gov (United States)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  11. Fuel behavior comparison for a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gh. [Institute for Nuclear Research (ICN), 1, Campului Street, P.O. Box 78, 0300 Mioveni, Pitesti (Romania)]. E-mail: joenegut@yahoo.com; Mladin, M. [Institute for Nuclear Research (ICN), 1, Campului Street, P.O. Box 78, 0300 Mioveni, Pitesti (Romania); Prisecaru, I. [University Politehnica Bucharest (Romania); Danila, N. [University Politehnica Bucharest (Romania)

    2006-06-30

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  12. Research nuclear reactor RA - Annual Report 1991

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Construction of some support elements is almost finished by the local staff. The Institute has undertaken this activity in order to speed up the ending of the project. If all the planned instrumentation would not arrive until the end of March 1992, it would not be possible to start the RA reactor testing operation in the first part of 1993, as previously planned. In 1991, 53 staff members took part in the activities during 1991, which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection

  13. Experimental investigations of graphite corrosion and aerosol formation during air ingress into the core of a high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    A High Temperature Reactor can be designed to remove the decay heat without using any active systems. For most accident scenarios a release of radioactive fission products can be excluded by design. However, during operation of a HTR some accidents are principally possible, which can result in a release of fission products out of the fuel elements and of the reactor system. One of these accidents is a hypothetical massive air ingress into the hot graphite reactor core. After a pressure drop caused by leakages in the primary circuit a gas mass flow may be able to stream through the core according to free natural convection leading to a corrosion of the graphite fuel elements and reflector structures. With the VELUNA-experiment a testing device was installed, which allows to investigate the corrosion process on parts of a reactor core under real accident conditions. With regard to the experimental results already existing equations to calculate the chemical reaction rate in a pebble bed were modified and the applicability was demonstrated. These equations consider the chemical reaction in the porous graphite as well as diffusion processes to the graphite surface. Equivalent correlations were developed for different flow geometries and for the graphite material of the bottom reflector. The corrosion process forms an aerosol, which consists of graphite particles and a reaction gas phase. The formatted aerosol was characterized concerning its chemical and physical properties. Because the aerosol particles can support the release of fission products, measurements of aerosol parameters like particle mass concentration and particle size distribution provide important information to estimate the radiologic consequences of such an hypothetical air ingress accident. (orig.)

  14. Proceedings of the Conference on research reactors application in Yugoslavia

    International Nuclear Information System (INIS)

    The Conference on research reactors operation was organised on the occasion of 20 anniversary of the RB zero power reactor start-up. The presentations showed that research reactors in Yugoslavia, RB, RA and TRIGA had an important role in development of nuclear sciences and technology in Yugoslavia. The reactors were applied in non-destructive testing of materials and fuel elements, development of reactor noise techniques, safety analyses, reactor control methods, neutron activation analysis, neutron radiography, dosimetry, isotope production, etc

  15. Comprehensive thermal hydraulics research of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: high temperature gas reactor fuels behaviour, high temperature materials qualification, design methods development and validation, hydrogen production technologies energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  16. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    The technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  17. Fast reactor systems for deep sea research

    International Nuclear Information System (INIS)

    Fast reactor (FR) systems have been studied as power units for unmanned bases and research submersibles to monitor various phenomena and as a thermal source for the unmanned base to feed useful microorganisms in the deep sea region. The systems, which are set in pressure hulls, comprise of the FR's and secondary gas loops. Concepts and arrangements of the systems are presented. (author)

  18. Safety status of Russian research reactors

    International Nuclear Information System (INIS)

    Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

  19. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories. AECL owns and operates the multi-purpose research reactor that serves as the primary R and D facility for supporting the CANDU business. The reactor is also a major producer of the world's medical radioisotopes. Since NRU was started up in 1957, it has operated in a consistent and safe manner with an overall annual capacity factor of approximately 80 %. The demands on the operation to perform experiments and produce radioisotopes were increased significantly when the NRX (National Research Experimental) shut down in 1992. Radioisotope customers demand an uninterrupted supply of short-lived radioisotopes e g Molybdenum-99, while experimental researchers require frequent shutdowns to accommodate fuel and materials programs. A two year systematic review and assessment of NRU to determine the condition and state of the facility was completed in 1991. This engineering assessment was complemented by safety analyses which focused on systems and components critical to safety. Reactor aging, obsolescence, current codes, and hazards vulnerability (especially, seismic) were emphasized during the analyses. This initial assessment concluded that the overall condition of NRU was good and there was no undue risk to the public or environment with the present operation. In addition, seven major upgrades were identified to enhance reactor safety to satisfy modern standards. In 1992, the AECL executive approved the Upgrades Project. Implementation of the seven upgrades were then included in the Facility Authorization document that defines the limiting conditions for safe operation with the Chalk River site license. The Atomic Energy Control Board would approve and license the upgrades under the change control provisions of the FA. Each upgrade and/or assessment recommendation (minor modification) had to be implemented without adversely affecting the current

  20. Proceedings of the European Research Reactor Conference - RRFM 2012 Transactions

    International Nuclear Information System (INIS)

    In 2012 RRFM, the European Research Reactor Conference will be jointly organised with IGORR, the International Group Operating Research Reactors. This will allow offering engineers and specialised nuclear researchers the chance to focus on the latest technological developments in the field of nuclear research reactors. The conference programme will revolve around a series of Plenary Sessions dedicated to the latest global developments with regards to research reactor technology and management systems, parallel sessions that focused on specific research projects and initiatives. (authors)

  1. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

    The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

  2. Using deterministic methods for research reactor studies

    International Nuclear Information System (INIS)

    As an alternative to prohibitive Monte Carlo simulations, deterministic methods can be used to simulate research reactors. Using various microscopic cross section libraries currently available in Canada, flux distributions were obtained from DRAGON cell and supercell transport calculations. Then, homogenization/condensation is done to produce few-group nuclear properties, and diffusion calculations were performed using DONJON core models. In this paper, the multigroup modular environment of the code DONJON is presented, and the various steps required in the modelling of SLOWPOKE hexagonal cores are described. Numerical simulations are also compared with experimental data available for the EPM Slowpoke reactor. (author)

  3. Assessment of a RELAP5 model for the IPR-R1 Triga research reactor

    International Nuclear Information System (INIS)

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this way, as a contribution to the assessment of RELAP5/3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed by a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open-pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data and also calculation data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code were considered in the process of the model validation. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual reactor behavior in good agreement with the available data. (author)

  4. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  5. Release from control of inactive material from decommissioning the ASTRA research reactor

    International Nuclear Information System (INIS)

    The Austrian Research Centers Seibersdorf have been operating a 10 MW ASTRA research reactor from 1960 until 1999. After that date, a submission of the intention to decommission the reactor has been provided to the Competent Authorities. After completion of an Environmental Impact Study by the Competent Authorities and modification of the Permissions for Site Use, the reactor finally entered the decommissioning phase in 2003. Inactive materials from the decommissioning site are expected to be released from control. The procedure for such a release from control agreed upon between the Competent Authorities and ARC Seibersdorf involves a four-step measurement, verification, and certification process detailed in this paper. By September 2003, this four-step procedure has been completed for 16500 kg of steel re-enforced concrete and for 5500 kg of other materials; the release from control of 3000 kg of paraffin and 10000 kg of graphite from the thermal column are planned for the near future. (author)

  6. Radio Nuclides Release in Research Reactors

    International Nuclear Information System (INIS)

    One of the major topic in nuclear safety is the quantitative evaluation of the radionuclides source term in nuclear reactors under routine and accidental conditions. The present study considers the release paths from fuel to coolant during normal and accidental situations of research reactors. Equivalent full power days approach, has been adopted for implementing reactor operating history in the calculations. Origin II code , recoil and Knock out phenomena, experimental correlations, and mathematical models have been employed in determining source term in fuel, releases to fuel clad interface, release from clad to coolant, and concentration in coolant. Different volatile fission products have been manipulated as: Br-83, Kr-85, I-129, I-131, I-133, Xe-133, Xe-135, Cs-137, Te-127, Te-131 m, Tc-99, Tc 99 m, Mo-99, Sr-90, Ru-106. Normal operation and accidental situation have been studied. The results have been verified against published data during normal operating conditions, it showed a good agreement

  7. Pakistan research reactor-1 and its upgradation

    International Nuclear Information System (INIS)

    In this article the author describes the procedure of renovation and upgradation of a swimming pool type Pakistan Research Reactor-1 (PARR-1) installed at PINSTECH. The reactor originally designed for a thermal power of 5 MW using highly enriched uranium as has been upgraded 10 MW with low enriched uranium as fuel. All the required safety precaution has been also modified with the new requirements. The cooling system of PARR-1 was modified to meet the requirements of upgraded power of 10 MW. In order to ensure safety for upgraded PARR-1 and to bring the reactor the current safety standards, some additional safety systems have been provided. An emergency core cooling system ECCS has been installed to remove core decay heat in case of loss of coolant accident (LOCA). (A.B.)

  8. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

  9. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1980 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)

  10. Analysis of the fast transients in the research reactor MARIA

    International Nuclear Information System (INIS)

    The analysis of the physical process in the research reactor MARIA during fast reactivity excursions is presented. The mathematical model of dynamics has been developed and its parameters as well as the temperature coefficients of reactivity were analysed. The analysis has shown that the change in the coolant density and fuel expansion are the most dominant temperature effects of reactivity. These effects differ by nearly two orders of magnitude from the reactivity effect due to fuel temperature changes, while the influence of the remaining materials of the core, i.e. beryllium and graphite, is negligible. Weak fuel temperature feedback is caused chiefly by the high enrichment of U-235, which indicates that the role of the Doppler effect significant for fast transients is practically unimportant. On the basis of the dynamics model, the code TOTEM for the CYBER-73 digital computer was written in FORTRAN-EXTENDED. The user's manual with input preparation and output explanation is enclosed. The code is intended for analysis of the reactor accidents caused by the rapidly increased reactivity. The following calculations can be carried out: axial and radial temperature distribution in the fuel channel with mean power; feed-back reactivity; power transients. The code was verified experimentally. A comparison of the measured and calculated feed-back reactivity and temperatures has shown good agreement. The space effects were investigated using signals from two neutron detectors, cobalt self-powered detector (SPD-Co) and ionization chamber located at a different distance from the perturbation point. All considered accidents, excluding the stepwise insertion the reactivity of 2$ (on the assumption that the protection system would respond properly) do not peril the safety of the reactor. 43 refs., 39 figs., 7 tabs. (author)

  11. Research on graphite powders used for HTR-PM fuel elements

    Institute of Scientific and Technical Information of China (English)

    ZHAO Hongsheng; LIANG Tongxiang; ZHANG Jie; LI Ziqiang; TANG Chunhe

    2006-01-01

    Different batches of natural graphite powders and electrographite powders were characterized by impurity, degree of graphitization, particle size distribution, specific surface area, and shape characteristics. The graphite balls consist of proper mix-ratio of natural graphite, electrographite and phenolic resin were manufactured and characterized by thermal conductivity, anisotropy of thermal expansion, crush strength, and drop strength. Results show that some types of graphite powders possess very high purity, degree of graphitization, and sound size distribution and apparent density, which can serve for matrix graphite of HTR-PM. The graphite balls manufactured with reasonable mix-ratio of graphite powders and process method show very good properties. It is indicated that the properties of graphite balls can meet the design criterion of HTR-PM. We can provide a powerful candidate material for the future manufacture of HTR-PM fuel elements.

  12. Nuclear reactors for research and radioisotope production in Argentina

    International Nuclear Information System (INIS)

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981

  13. Thermal Insulation Properties Research of the Composite Material "Water Glass - Graphite Microparticles"

    Science.gov (United States)

    Gostev, V. A.; Pitukhin, E. A.; Ustinov, A. S.; Shelestov, A. S.

    2016-04-01

    Research results for the composite material (CM) "water glass - graphite microparticles" with high thermal stability and thermal insulation properties are given. A composition is proposed consisting of graphite (42 % by weight), water glass Na2O(SiO2)n (50% by weight) and the hardener - sodium silicofluoride Na2SiF6 (8% by weight). Processing technology of such composition is suggested. Experimental samples of the CM with filler particles (graphite) of a few microns in size were obtained. This is confirmed by a study of samples using X-ray diffraction analysis and electron microscopy. The qualitative and quantitative phase analysis of the CM structure was done. Values of limit load causing destruction of the CM were identified. The character of the rupture surface was detected. Numerical values of the specific heat and thermal conductivity were defined. Dependence of the specific heat capacity and thermal conductivity on temperature during monotonic heating was obtained experimentally. Studies have confirmed the increased thermal insulation properties of the proposed composition. The CM with such properties can be recommended as a coating designed to reduce heat losses and resistant to high temperatures. Due to accessibility and low cost of its components the proposed material can be produced on an industrial scale.

  14. Methods And Results Of Reconstruction Of Noble Gas Releases From The Stacks Of The Mayak PA Graphite Reactors Over The Whole Period Of Their Operation

    International Nuclear Information System (INIS)

    Brief analysis of design features and operational modes of Mayak PA industrial graphite-uranium reactors (PUGRs) is given. The above mentioned Mayak PA PUGRs determined the rates of releases of radioactive noble gases (RNG) from activation (41Ar) and fission (isotopes of Krypton and Xenon) through the vent stack of the reactor. Information is given on methods and results of experimental determination of RNG atmospheric releases for the period starting from 1965 till PUGRs decommissioning in 1987-1990. A calculation method for reconstruction of radioactive noble gas releases is proposed and justified. The results of reconstruction are given. It is shown that maximum rates of RNG releases from PUGRs high stacks were observed in the 1950s, when ordinary atmospheric air was used as a cover gas for the reactor graphite stacks and gas purification systems (flow-type gas holders) had not been installed yet.

  15. METHODS AND RESULTS OF RECONSTRUCTION OF NOBLE GAS RELEASES FROM THE STACKS OF THE MAYAK PA GRAPHITE REACTORS OVER THE WHOLE PERIOD OF THEIR OPERATION

    Energy Technology Data Exchange (ETDEWEB)

    Glagolenko, Y. V.; Drozhko, Evgeniy G.; Mokrov, Y.; Pyatin, N. P.; Rovny, Sergey I.; Anspaugh, L. R.; Napier, Bruce A.

    2008-06-01

    Brief analysis of design features and operational modes of Mayak PA industrial graphite-uranium reactors (PUGRs) is given. The above mentioned Mayak PA PUGRs determined the rates of releases of radioactive noble gases (RNG) from activation (41Ar) and fission (isotopes of Krypton and Xenon) through the vent stack of the reactor. Information is given on methods and results of experimental determination of RNG atmospheric releases for the period starting from 1965 till PUGRs decommissioning in 1987-1990. A calculation method for reconstruction of radioactive noble gas releases is proposed and justified. The results of reconstruction are given. It is shown that maximum rates of RNG releases from PUGRs high stacks were observed in the 1950s, when ordinary atmospheric air was used as a cover gas for the reactor graphite stacks and gas purification systems (flow-type gas holders) had not been installed yet.

  16. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO.

  17. Description of the french graphite reactor and of the experiments performed in 1956

    International Nuclear Information System (INIS)

    This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author)

  18. Experience of operating and decommissioning of the research IRT-M reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of IRT-type was being operated during 25 years at the Institute of Power Engineering Problems /Academy of Sciences of Belarus (the former Institute of Nuclear Power Engineering/Academy of Sciences of Byelorussian SSR). The reactor with the designed power of 2000 kW was put into operation in April 1962 . With the same level of power the reactor was under operation up to 1971 with using rod fuel elements of 10%- enrichment as to U-235 as fuel and a side graphite reflector. As the research works have been developed, it became clear that comparatively small power of the reactor and, as the consequence of this, low level of neutron flux in experimental channels were insufficient for further development of the investigations. Besides, the reactor core with downwards upwards circulation of the coolant has led to vibrations of fuel assemblies, suspended in up going flow of the coolant. Hood-baffle plate of water flow and assembly-holder above the core, which are necessary in the scheme of cooling given, restricted the accessibility to the core for carrying out experiments there. In 1971 the modernization of the reactor was done directed to the increasing of its power up to 4000 kW and improving its design but with preserving the main designed principles peculiar to a basin-type reactor. Experience of Operating and Decommissioning of the IRT-M Reactor are presented. (author)

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

    International Nuclear Information System (INIS)

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

  20. New developments in transportation for research reactors

    International Nuclear Information System (INIS)

    For more than 30 years, Transnucleaire has been performing safely a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied numerous packagings for all types of nuclear fuel cycle radioactive materials: for front-end and back-end products and for power and research reactors. Since the last meeting held in Bruges, Transnucleaire has been continuously involved in transportation activities for fresh and irradiated materials for research reactors. We are pleased to take the opportunity in this meeting to share with reactor operators, official bodies and other partners, the on-going developments in transportation and associated services. Special attention will be paid to the starting of transports of MTR spent fuel elements to the La Hague reprocessing plant where COGEMA offers reprocessing services on a long-term basis to reactors operators. Detailed information is provided on regulatory issues, which may affect transport activities: evolution of the regulations, real experiences of recent transportation and development of new packaging designs. Options and solutions will be proposed by Transnucleaire to improve the situation for continuation of national and international transports at an acceptable price whilst maintaining an ultimate level of safety (author)

  1. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  2. Radionuclide release from research reactor spent fuel

    International Nuclear Information System (INIS)

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO2-fuel (LWR fuel, enrichment in 235U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl2-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAlx-Al and U3Si2-Al) was studied in 400 mL MgCl2-rich salt brine in the presence of Fe2+ under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH)3(s) and Eu(OH)3(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu(IV) oxyhydroxides species due to radiolytic effects cannot completely be

  3. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  4. Research reactor fuel management in the Czech Republic

    International Nuclear Information System (INIS)

    Fuel management of the Czech research reactors is described. There are three research reactors in the Czech Republic: LVR-15 and LR-0 operated by the Nuclear Research Institute Rez plc, VR-1 operated by the Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering in Prague, and SR-0 reactor of SKODA JS plc which is under decommissioning now. The paper describes the major features of the Czech research reactors, types of fuels used in them, and the spent fuel management principles. The participation of the LVR-15 and VR-1 reactors in the RERTR international programme (Reduced Enrichment for Research and Test Reactors) is also highlighted. (author)

  5. Current activities at the MIT Research Reactor

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a MW nuclear research reactor that is owned and operated by the Massachusetts Institute of Technology to further its educational and research goals at both the undergraduate and graduate level. The reactor first achieved criticality in 1958. It was largely rebuilt in 1973/1974 by MIT staff and students, and its current license expires in August 1999. The current facility, which is designated as the MITR-H, uses a compact core with finned, aluminum-clad, plate-type fuel that is cooled and moderated by light water and reflected by heavy water. The reactor core can hold twenty-seven fuel elements. However, the normal configuration is twenty-four elements. A maximum of four fuel elements can be replaced with in-core experimental facilities. A unique feature of the MITR-II's design is that fixed absorber plates can be inserted in the upper half of the core. These cause the flux to peak in the lower half which benefits experimenters and also facilitates a fuel strategy that involves inversion of fuel elements midway through their life cycle. The MITR-II currently operates continuously for four weeks followed by shutdown of a few days for maintenance. This paper provides an overview of current activities at the MITR including preparations for re-licensing. The status of an on-going Phase-I clinical trial of boron neutron capture therapy for both glioblastoma multiforme and metastatic melanoma is described as well as the design of a fission converter facility for BNCT. Environmental research using neutron activation analysis is summarized as well as in-pile research focussed on LWR water chemistry and structural materials. (author)

  6. EURATOM Research Framework Programme on Reactor Systems

    International Nuclear Information System (INIS)

    The activities of the European Commission (EC) in the field of nuclear energy are governed by the Treaty establishing the European Atomic Energy Community (EURATOM). The research activities of the European Union (EU) are designed as multi-annual Framework Programmes (FP). The EURATOM 6. Framework Programme (EURATOM FP -6), covering the period 2002-2006, is funded with a budget of 1, 230 million Euros and managed by the European Commission. Beyond the general strategic goal of the EURATOM Framework Programmes to help exploit the potential of nuclear energy, in a safe and sustainable manner, FP -6 is designed to contribute also to the development of the 'European Research Area' (ERA), a concept described in the Commission's Communication COM(2000)6, of January 2000. Moreover EURATOM FP-6 contributes to the creation of the conditions for sharing the same nuclear safety culture throughout the EU-25 and the Candidate Countries, fostering the acceptance of nuclear power as an element of the energy mix. This paper gives an overview of the research activities undertaken through EURATOM FP-6 in the area of Reactor Systems, covering the safety of present reactors, the development of future safe reactors, and the needs in terms of research infrastructures and education and training. The actions under FP-6 are presented in their continuity of actions under FP-5. The perspectives under FP -7 are also provided. Other parts of the EURATOM FP, covering Waste Handling and Radiation Protection, as well as Fusion Energy, are not detailed in this paper. (authors)

  7. An Overview of Ageing Management and Refurbishment of Research Reactors at Trombay

    International Nuclear Information System (INIS)

    Three nuclear research reactors have been in operation at Bhabha Atomic Research Centre, Mumbai, India. India has a rich experience of about 120 research reactor operating years including ageing management. A well structured programme is in force for plant life management, refurbishment and upgrading reactors in operation. Apsara, commissioned in August 1956, was the first research reactor. Apsara is a 1 MWth swimming pool type of reactor with a movable core loaded with enriched uranium fuel and immersed in demineralized light water pool, which serves as coolant, moderator and reflector besides providing radiation shielding. Apsara was shut down during May 2009 for partial decommissioning and upgrading to a 2 MW reactor with several safety upgrades, e.g. a LEU based reactor core with higher neutron flux, a new reactor building meeting seismic qualification criteria and two independent shutdown devices. Cirus, a 40 MWth tank type reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; has been in operation since 1960. After about three decade of operation, the availability factor started declining mainly due to outage of equipment exhibiting signs of ageing. After ageing studies and performance review, refurbishment requirements were identified. A programme for refurbishment was drawn that included safety upgrades like civil repairs to the emergency storage reservoir to meet seismic qualification criteria and a new iodine removal system for better efficiency. The reactor was shut down during 1997 for execution of this refurbishment programme. After completion of refurbishment, the reactor was brought back into operation during 2003. It has completed about seven years of safe operation after refurbishment with a significant increase in availability factor from 70% to about 90%. The reactor was permanently shut down during December 2010. The reactor core was unloaded, and

  8. Proceedings of the Oak Ridge National Laboratory/Brookhaven National Laboratory workshop on neutron scattering instrumentation at high-flux reactors

    International Nuclear Information System (INIS)

    For the first three decades following World War II, the US, which pioneered the field of neutron scattering research, enjoyed uncontested leadership in the field. By the mid-1970's, other countries, most notably through the West European consortium at Institut Laue-Langevin (ILL) in Grenoble, France, had begun funding neutron scattering on a scale unmatched in this country. By the early 1980's, observers charged with defining US scientific priorities began to stress the need for upgrading and expansion of US research reactor facilities. The conceptual design of the ANS facility is now well under way, and line-item funding for more advanced design is being sought for FY 1992. This should lead to a construction request in FY 1994 and start-up in FY 1999, assuming an optimal funding profile. While it may be too early to finalize designs for instruments whose construction is nearly a decade removed, it is imperative that we begin to develop the necessary concepts to ensure state-of-the-art instrumentation for the ANS. It is in this context that this Instrumentation Workshop was planned. The workshop touched upon many ideas that must be considered for the ANS, and as anticipated, several of the discussions and findings were relevant to the planning of the HFBR Upgrade. In addition, this report recognizes numerous opportunities for further breakthroughs on neutron instrumentation in areas such as improved detection schemes (including better tailored scintillation materials and image plates, and increased speed in both detection and data handling), in-beam monitors, transmission white beam polarizers, multilayers and supermirrors, and more. Each individual report has been cataloged separately

  9. Papers concerning fuel particles, fuel elements and graphitic materials for high temperature reactors presented at the 1980 annual conference Kerntechnik

    International Nuclear Information System (INIS)

    This report is a compilation of the papers presented by staff of the Institute for Reactor Materials, KFA-Juelich, at the 1980 annual conference Rekatortechnik, held in Berlin, 25-27th March 1980. In some cases, there were co-authors from other organisations. Where possible the manuscripts of the presentations have been reproduced, as well as the display cards shown during the poster session on the conference. In the presentations, the questions of the characterization of fuel particles and the retention of fission products are dealt with, and special attention is given to fission product release at very high temperatures. One presentation deals with the disposal of fuel elements from the AVR. Another report presents the results of radiation experiments on the standard matrix material A3-3 of the THTR fuel elements; the changes in dimensions, creep coefficient and thermal conductivity were measured as functions of the fluence and the radiation temperature. Interim results obtained from long-term radiation experiments on reflector graphites for high and very high flux reactors are presented, and models for the calculation of dimensional changes in components subjected to fluctuating temperatures from data obtained in isothermal tests are discussed. (orig.)

  10. Decommissioning of uranium-graphite nuclear reactors and radioactive waste handling

    International Nuclear Information System (INIS)

    In accordance with the agreements between the Russian and US Governments on completion of weapons plutonium production, ten powerful industrial reactors were unloaded and decommissioned in the Russian Federation. Three such reactors, I1, E2 and ADE3, were decommissioned after 30 years of operation. The decommissioning concept started in the middle 1990s, comprising the dismantling of low-activity structures, sealing of all reactor outlets, and filling of all reactor spaces with special compounds of concrete and betonite. All these arrangements ensured multi-level protection with a number of safety barriers between the reactor and the environment. As a result, the suggested decontamination and decommissioning concept allowed qualification of the structure under the IAEA's Stage 2 and ensured its safety for over 300 years.The presence of radiation and restricted access to the reactor components led to the development of remote controlled equipment and for electrical contact or mechanical cutting methods, for welding and remote sampling of radioactive metal and other items. The paper also discusses the cost of the work that has been performed, personnel exposures, the quantity of filling materials consumed for the support structures, as well as the handling of radioactive wastes (i.e. reactor products).The ways they were unloaded, their selection, how they were contained, their transport and termination are described. Particular attention is paid to the use of special compounds with betonite ingredients as a protection against radiation, which ensures multi-layered protection for radioactive waste on the basis of the 'matrix-isolation-coat' principle. This allows the radioactive waste to be stored at a decommissioning site. (author)

  11. Research Reactors: Decommissioning of a Small Reactor (BR3 Reactor, Belgium). Appendix III

    International Nuclear Information System (INIS)

    Research reactors are nuclear reactors that serve primarily as source of neutrons. They are less complex than power reactors and operate at lower temperatures. Research reactors need far less fuel, and far less fission products build up as the fuel is used. On the other hand, their fuel requires more highly enriched uranium, typically up to 20% 235U. More than 650 research reactors worldwide have been built or are under construction or in a planning phase; of which more than 350 have been shut down and partly or wholly decommissioned. Experience has shown that decommissioning can be undertaken in line with safety standards aimed at protecting human beings or the environment from harm, provided that decommissioning activities are undertaken in accordance with a properly formulated plan. The potential or actual radiological hazards associated with reactors may require the application of special techniques and procedures during decommissioning. The decommissioning of the BR3 reactor in Mol, Belgium, Belgian nuclear research centre SCK•CEN, provides an example of current good practice in decommissioning research reactors.13 Since 1991, the organization’s statutory mission gives priority to research on problems of societal concern such as the safety of nuclear installations, radiation protection, safe treatment and disposal of radioactive waste, fighting against uncontrolled proliferation of fissile materials, and education and training. BR3 was the first European pressurized water reactor (PWR) power plant and was put into service in 1962. It was in that industrial context that the BR3 has played its role as a demonstration unit for the development and improvement of decommissioning related techniques. While the BR3 power level was low (40 MW(th), 10.5 MW(e) net), it contains all the features of commercial PWR power plants. The reactor was used at the beginning of its lifetime as a training facility for future nuclear power plant operators. Later, it was also used

  12. A New Generation of Research Reactors Fuelled with LEU

    International Nuclear Information System (INIS)

    A number of countries have recently shown interest in new research reactors. In response to such willingness to develop nuclear technologies, we have prepared technical proposals on typical research reactors (RR) which will be built as part of nuclear research centres (NRC) according to base design principles. The requirements for such research reactors are defined to represent their competitive service parameters, including capabilities to support a wide spectrum of studies in various areas of theoretical and applied researches. Analysis of the current and projected uses of research reactors and assessment of the external market demands have prompted two design options of a pool-type reactor at a nuclear research centre, namely, a small (up to 0.5 MW) reactor with natural coolant circulation through its core and a reactor with forced coolant circulation scaled up to 10-15 MW. The research reactors under development will run with commercially available and well-proven fuel of low enrichment. (author)

  13. The current status of utilization of research reactors in China

    International Nuclear Information System (INIS)

    Seminars on utilization of research reactors were held to enhance experience exchanging among institutes and universities in China. The status of CARR (China Advanced Research Reactor) project is briefly described. The progress in BNCT program in China is introduced. (author)

  14. Safety Features of the Replacement Research Reactor

    International Nuclear Information System (INIS)

    The paper presents a general description of the development and application of basic safety criteria and the implementation of specific safety features in the design of the 20 MW pool-type research reactor currently being built by INVAP for the Australian Nuclear Science and Technology Organisation (ANSTO). A summary of the results of the preliminary deterministic safety analysis and the probabilistic safety assessment prepared by INVAP on ANSTO's behalf are presented as part of demonstrating the robustness of the design to the wide range of postulated initiating events considered. The paper also briefly describes the licensing process with respect to the way in which the licensing and regulatory regime within Australia influenced the design of the replacement research reactor (RRR). In particular, the reasoning for safety design features that have been incorporated as a result of the specific requirements of ANSTO and the Australian regulator is described. (author)

  15. Decommissioning Experience: Apsara Research Reactor, India

    International Nuclear Information System (INIS)

    Full text: In India, at the Bhabha Atomic Research Centre, a 1 MW(th) pool type research reactor called Apsara was built in 1956 and shut down in 2009. The reactor fuel and internals were removed, leaving the pool available for draining and decontamination. The pool was drained progressively while monitoring for hot spots. Additional material and debris at the bottom were removed. The lining was cleaned by water jetting using detergents. In summary, the defuelling and partial decommissioning were successfully completed in around six months, with a total dose consumption of 23.5 man mSv (approximately 10% of budget). The generation of waste amounted to a solid waste volume of around 20 m3 (low level) and a liquid waste volume of 280 m3 (low level). A detailed description of achievements and plans for the Apsara decommissioning is given. (author)

  16. Australia's new high performance research reactor

    International Nuclear Information System (INIS)

    A contract for the design and construction of the Replacement Research Reactor was signed in July 2000 between ANSTO and INVAP from Argentina. Since then the detailed design has been completed, a construction authorization has been obtained, and construction has commenced. The reactor design embodies modern safety thinking together with innovative solutions to ensure a highly safe and reliable plant. Also significant effort has been placed on providing the facility with diverse and ample facilities to maximize its use for irradiating material for radioisotope production as well as providing high neutron fluxes for neutron beam research. The project management organization and planing is commensurate with the complexity of the project and the number of players involved. (author)

  17. Pakistan research reactor and its utilization

    International Nuclear Information System (INIS)

    The 5 MW enriched uranium fuelled, light water moderated and cooled Pakistan Research reactor became critical on 21st December, 1965 and was taken to full power on 22nd June, 1966. Since then is has been operated for about 23000 hours till 30th June, 1983 without any major break down. It has been used for the studies of neutron cross-sections, nuclear structure, fission physics, structure of material, radiation damage in crystals and semiconductors, studies of geological, biological and environmental samples by neutron activation techniques, radioisotope production, neutron radiography and for training of scientists, engineers and technicians. In the paper we have described briefly the facility of Pakistan Research Reactor and the major work carried around it during the last decade. (author)

  18. Developing the fuel for research reactors

    International Nuclear Information System (INIS)

    A review of papers dealing with the possibility of research reactor adaptation to moderately and slightly enriched fuel with the 235U content of 45 and 20%, respectively, is presented. The main peculiarities and results of investigations carried out in two main directions, are under consideration: the increase of specific uranium content in traditional fuels (UAlsub(x)-Al, U3O8-Al, U,ZrHsub(x)) by means of improvements in technology and production (USA, FRG and France); the development of new highly dense kinds of fuel, such as U3Si, U3Si-Al, UO2 (USA, France). A conclusion is drawn that the research reactor fuel enrichment may be decreased

  19. A new fuel for research reactors

    International Nuclear Information System (INIS)

    The Replacement Research Reactor (RRR) to be constructed at Lucas Heights will use fuel containing low enriched uranium (LEU), 235U, whereas its predecessor HIFAR operates with fuel fabricated from high-enriched uranium (HEU). The fuel will be based on uranium silicide (U3Si2) with a density of 4.8 g U/cm3. This fuel has been qualified and in use in 20 research reactors worldwide for over 12 years A brief description is given of the metallurgy, behaviour under irradiation, and fabrication methods, all of which are well-understood Progress on development of new, higher density LEU fuel based on uranium molybdenum alloys is also described and the implications for the RRR discussed briefly

  20. Reactor training simulator for the Replacement Research Reactor (RRR)

    International Nuclear Information System (INIS)

    The main features of the ANSTO Replacement Research Reactor (RRR) Reactor Training Simulator (RTS) are presented.The RTS is a full-scope and partial replica simulator.Its scope includes a complete set of plant normal evolutions and malfunctions obtained from the plant design basis accidents list.All the systems necessary to implement the operating procedures associated to these transients are included.Within these systems both the variables connected to the plant SCADA and the local variables are modelled, leading to several thousands input-output variables in the plant mathematical model (PMM).The trainee interacts with the same plant SCADA, a Foxboro I/A Series system.Control room hardware is emulated through graphical displays with touch-screen.The main system models were tested against RELAP outputs.The RTS includes several modules: a model manager (MM) that encapsulates the plant mathematical model; a simulator human machine interface, where the trainee interacts with the plant SCADA; and an instructor console (IC), where the instructor commands the simulation.The PMM is built using Matlab-Simulink with specific libraries of components designed to facilitate the development of the nuclear, hydraulic, ventilation and electrical plant systems models

  1. Chemistry research and chemical techniques based on research reactors

    International Nuclear Information System (INIS)

    Chemistry has occupied an important position historically in the sciences associated with nuclear reactors and it continues to play a prominent role in reactor-based research investigations. This Panel of prominent scientists in the field was convened by the International Atomic Energy Agency (IAEA) to assess the present state of such chemistry research for the information of its Member States and others interested in the subject. There are two ways in which chemistry is associated with nuclear reactors: (a) general applications to many scientific fields in which chemical techniques are involved as essential service functions; and (b) specific applications of reactor facilities to the solution of chemical problems themselves. Twenty years of basic research with nuclear reactors have demonstrated a very widespread, and still increasing, demand for radioisotopes and isotopically-labelled molecules in all fields of the physical and biological sciences. Similarly, the determination of the elemental composition of a material through the analytical technique of activation analysis can be applied throughout experimental science. Refs, figs and tabs

  2. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  3. Decommissioning of a 5 MW research reactor

    International Nuclear Information System (INIS)

    The complete decommissioning of a research reactor is described. Planning and execution of all activities, including schedules, budgets, waste management, health physics and subcontracted operations are presented. Flexibility in operations was obtained by using the operating staff as the decommissioning progressed. Totals for waste shipments and costs are given. Final site conditions are presented along with a description of the subsequent use of the facility. (author)

  4. Hydrogen problems in reactor safety research

    International Nuclear Information System (INIS)

    The BMFT and BMI have initiated a workshop 'Hydrogen Problems in Reactor Safety Research' that took place October 3./4., 1983. The objective of this workshop was to present the state of the art in the main areas - Hydrogen-Production - Hydrogen-Distribution - Hydrogen-Ignition - Hydrogen-Burning and Containment Behaviour - Mitigation Measures. The lectures on the different areas are compiled. The most important results of the final discussion are summarized as well. (orig.)

  5. Defuelling of the UTR-300 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scott, R.D.; Banford, H.M.; East, B.W. [Scottish Universities Research and Reactor Centre, Glasgow (United Kingdom); Ord, M.A.; Gaffka, A.P. [AEA Technology, Harwell, Didcot, Oxfordshire OX11 0RA (United Kingdom)

    1997-12-01

    A description is given of the movement of fuel elements from the core of the UTR-300 research reactor to the UNIFETCH flask, which is normally loaded under water, through a specially designed shielding arrangement which permits a dry transfer. The regulatory requirements and the safety case are summarised along with the predicted and measured doses to operators. The task was successfully completed to a tight time schedule with recorded doses which were well within the allocated dose budget. (orig.) 3 refs.

  6. Safety review, assessment and inspection on research reactors, experimental reactors and nuclear heating reactors

    International Nuclear Information System (INIS)

    The NNSA and its regional office step further strengthened the regulation on the safety of in-service research reactors in 1996. A lot of work has been done on the supervision of safe in rectifying the review and assessment of modified items, the review of operational documents, the treatment of accidents, the establishment of the system for operational experience feedback, daily and routine inspection on nuclear safety. The internal management of the operating organization on nuclear safety was further strengthened, nuclear safety culture was further enhanced, the promotion in nuclear safety and the safety situation for in-service research reactors were improved

  7. FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

    International Nuclear Information System (INIS)

    The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted

  8. Organization and management of operation of the research reactor MARIA

    International Nuclear Information System (INIS)

    The MARIA research reactor belongs to the Institute of Atomic Energy. The MARIA research reactor operation provides basing on the Atomic Law code and requirements of the State Nuclear Safety. Main task of the operation Department is the current MARIA reactor operation and relevant technological systems. The Head of the Reactor bears the direct responsibility for nuclear safety and radiological protection of the reactor plant. Service of reactor operation is accomplished by the Shift Groups. The cooperation with the reactor users is based on the principles defined by the Regulations of MARIA Reactor Operation. In the abnormal and emergency states the procedure is determined by 'Schedule of emergency procedure for the MARIA reactor plant'. Reactor has got valid and actual documents which are compulsory to all the persons being involved in operation and usage of reactor. (author)

  9. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    International Nuclear Information System (INIS)

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies

  10. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  11. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1978 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, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

  12. Research and Development of Fast Reactor Metal Fuel by CRIEPI

    International Nuclear Information System (INIS)

    CRIEPI has been developing metal fuel to demonstrate its applicability to commercialized fast reactors, promote its practical use and help secure a low-cost electricity supply in the future. The achievements of its research and development to date can be summarized as follows. Analysis by the computer code ALFUS suggested that FCMI would remain at a low level even at a higher peak burnup, >10∼15 at.%. The threshold temperature of the liquefaction at the fuel-cladding interface was estimated at 923 K when Pu/(U+Pu)>0.25, based on the ex-reactor experiments. The conservative evaluation of the linear-power-to-melting confirmed a sufficient margin to fuel slug melting. An irradiation test of metal fuel in Joyo was planned to obtain irradiation performance data at high cladding temperatures and high burnups. Fundamental research was also conducted on the diffusion behavior in U-Zr alloys and the thermodynamics of U-Pu base alloys. Engineering-scale injection casting tests with U-Zr alloys showed that fuel slugs consistent with the practical reactor core design and fuel fabrication process could be fabricated at a reasonable throughput. The U-Pu-Zr fuel slugs to be irradiated in Joyo were fabricated from the U-Pu alloy ingots prepared by the electrochemical reduction of MOX pellets. Selective reaction of the yttria coating and graphite crucible with Am and Pu was found after the injection casting of U-Pu-Zr fuel slug. The basic characteristics of MA-containing alloys were understood, such as the constituent miscibility and thermal conductivity. The MA-containing metal fuel pins were irradiated in the Phenix reactor, and post-irradiation examination thereof is underway in ITU. No significant effect of the MA addition to metal fuel on the irradiation performance has been found at this stage. Future works will be required, such as irradiation tests of the full-length and MA-containing metal fuel pins at high peak cladding temperatures up to high peak burnups. (author)

  13. Natural uranium-graphite system. Critial experiments on the G1 reactor

    International Nuclear Information System (INIS)

    A number of experiments have been performed during the start up period of the G1 (1956) and G2 (1958) reactors in Marcoule, both on their lattices and on different lattices (hollow rods, clusters, under moderated lattices). The first chapter gives a thorough description of the two reactors. The second chapter deals with buckling measurements, both absolute (flux plots) and relative by the method of progressive substitution. The experimental results are summarised in Table VI. The third chapter contains a number of other measurements performed on G1. (author)

  14. Radio-active pollution near natural uranium-graphite-gas reactors

    International Nuclear Information System (INIS)

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors)

  15. Materials research with neutron beams from a research reactor

    International Nuclear Information System (INIS)

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  16. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  17. 78 FR 58575 - Review of Experiments for Research Reactors

    Science.gov (United States)

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer...

  18. Progress activity of Thai Research Reactor in 2002

    International Nuclear Information System (INIS)

    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. TRR-1/M1 has been operated and utilized for various applications such as neutron activation analysis, radioisotope production, gem irradiation, neutron radiography and research works. To expand and promote the utilization of research reactor, the new 10 MW Research Reactor will be established in the Ongkarak Nuclear Research Center (ONRC) project and the project will be finished in the near future. (author)

  19. Research reactor status for future nuclear research in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel [Commissariat a l' Energie Atomique - CEA (France)

    2010-07-01

    During the 1950's and 60's, the European countries built several research reactors, partially to support their emerging nuclear-powered electricity programs. Now, over forty years later, the use and operation of these reactors have both widened and grown more specialized. The irradiation reactors test materials and fuels for power reactors, produce radio-isotopes for medicine, neutro-graphies, doping silicon, and other materials. The neutron beam reactors are crucial to science of matter and provide vital support to the development of nano-technologies. Other reactors are used for other specialized services such as teaching, safety tests, neutron physics measurements... The modifications to the operating uses and the ageing of the nuclear facilities have led to increasing closures year after year. Since last ENC, for example, we have seen, only in France, the closure of the training reactor Ulysse in 2007, the closure of the safety test dedicated reactor Phebus in 2008 and recently the Phenix reactor, last fast breeder in operation in the European Community, has been shut down after a set of 'end of life' technological and physical tests. For other research reactors, safety re-evaluations have had to take place, to enable extension of reactor life. However, in the current context of streamlining and reorganization, new European tools have emerged to optimally meet the changing demands for research. However the operation market of these reactors seems now increasing in all fields. For the neutron beams reactors (FRMII, ORPHEE, ILL, ISIS,..) the experimental needs are increasing years after years, especially for nano sciences and bio sciences new needs. The measurement of residual stress on manufactured materials is also more and more utilised. All these reactors have increasing utilizations, and their future seems promising. A new project project based on a neutron spallation is under definition in Sweden (ESSS: European Spallation Source

  20. The korea multi-purpose research reactor

    International Nuclear Information System (INIS)

    This paper presents and discusses background and status of the design of the 30MW Korea Multi-purpose Research Reactor(KMRR) which is planed to achieve its first criticality in December, 19992, at Daeduk site of the Korea Advanced Energy Research Institute (KAERI). KAERI playing the leading role in Korea's nuclear technology development takes the total responsibility for its design, construction and operation. Number of Korean nuclear industries are, also, actively participating in the project while making the most of their expertise in relevant areas. (Author)