WorldWideScience

Sample records for cabri reactor restarting

  1. The reactor Cabri

    International Nuclear Information System (INIS)

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

  2. New results from pulse tests in the CABRI reactor

    International Nuclear Information System (INIS)

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared

  3. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  4. CABRI Reactor: The fast neutron Hodoscope Calibration curves calculation with MORET

    Science.gov (United States)

    Bernard, Franck; Chevalier, Vincent; Venanzi, Damiano

    2014-06-01

    This poster presents the Hodoscope calibration curves calculation with 3D Monte Carlo code MORET. The fast neutron hodoscope is a facility of the CABRI research reactor at Cadarache (FRANCE). This hodoscope is designed to measure the fuel motion during a RIA in a pressurized water reactor. The fuel motion is measured by counting fast fission neutrons emerging from the test fuel placed in an experimental loop functioning like a Pressurized Water Reactor (T=300°C and P=155 bar), at the center of the CABRI core. The detection system of the hodoscope measures a signal which is a function of the fuel motion. The calibration curves allow then to convert the signal in a fuel mass. In order to calculate these curves, we have developed a method based on a Monte Carlo calculation code.

  5. Operational readiness reviews for restart of L reactor

    International Nuclear Information System (INIS)

    The L Reactor at the Savannah River Plant is being restarted after being in a standby status since 1968. Operational Readiness Reviews (ORRs) were conducted by DOE-SR and contractor personnel concurrent with the restart activity. This paper summarizes the ORR activity

  6. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig.

  7. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    International Nuclear Information System (INIS)

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig

  8. Safety Evaluation Report Restart of K-Reactor Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed.

  9. Safety Evaluation Report Restart of K-Reactor Savannah River Site

    International Nuclear Information System (INIS)

    In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed

  10. The French authority of nuclear safety (ASN) authorizes the restart-up of Cattenom's reactor 3

    International Nuclear Information System (INIS)

    On August 31, 2001, the French authority of nuclear safety (ASN) gave permission to Electricite de France (EdF) to restart the reactor no 3 of Cattenom's power plant. This reactor encountered important degradations of its fuel assemblies during its previous operating cycle which led to a level 1 incident on the INES scale. Thus the ASN has imposed to EdF a reinforced surveillance of the primary circuit of the reactor. This document brings together the different press releases, information notes, and rulings that have been written by the ASN about the defects that occurred on the fuel rods of Cattenom's reactor 3. (J.S.)

  11. Safety Evaluation Report: Restart of K-Reactor, Savannah River Site

    International Nuclear Information System (INIS)

    In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, ''Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues which need to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied a the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. This document supplements the April 1991 SER. The SER Supplement documents the resolution of several of the open items identified in the SER. Only those issues (sections) for which at least one open item identified in the SER has now been closed are addressed in this Supplement. Additionally, some SER sections had no open items identified. Therefore, this Supplement does not include all sections that were addressed in the SER. If there are any open items remaining to be resolved for the sections included in this Supplement, that is so identified at the end of the section. The resolution of those remaining open times, and all remaining open items for those SER sections not included in this first Supplement, will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed

  12. Recommendations for a restart of Molten Salt Reactor development

    International Nuclear Information System (INIS)

    The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 232U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

  13. Review of Savannah River Site K Reactor inservice inspection and testing restart program

    International Nuclear Information System (INIS)

    Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

  14. RIA tests in CABRI with MOX fuel

    International Nuclear Information System (INIS)

    Three MOX-fuel tests have been successfully performed within the framework of the CABRI REP-Na test program. From the experimental findings which are presently available, no evidence for thermal effects resulting from the heterogeneous nature of the fuel can be given. There are very clear hints however that fission gas effects are enhanced with regard to the behaviour of UO2 . The clad rupture observed in REP-Na 7 is of different nature than the failures observed in Cabri tests with UO2 fuel. Failures of UO2 fuel rods only occurred when the clad mechanical properties were severely affected by the presence of hydride blisters, while in REP-Na 7 a clear indication is made that the loading potential of the MOX fuel pellets was high enough to break a sound cladding. Concerning the transient fuel behaviour after reaching the critical heat-flux under reactor typical conditions (pressure, temperature and flow), no data base could be provided by the tests in the present sodium test loop (as for the UO2 fuel behaviour). The IPSN project to implement into the Cabri reactor a pressurised water loop which will allow to simulate the complete RIA accident sequence under PWR reactor typical conditions, aims at providing this missing data base. (author)

  15. Cabri e internet

    OpenAIRE

    Pérez, Antonio

    2001-01-01

    La aparición hace ya unos cuantos años del programa CabriGéomètre supuso para muchos profesores y profesoras la apertura de una ventana de esperanza en el camino de ver y de enseñar la geometría de una forma diferente. El éxito de la filosofía del programa radicaba en la idea de poder contar con una pizarra electrónica en la que construir objetos geométricos tan habituales como trazar rectas, segmentos, perpendiculares, ángulos, triángulos, circunferencias, cónicas... y medir en forma directa...

  16. Evaluation of the Steel Creek ecosystem in relation to the proposed restart of L reactor

    International Nuclear Information System (INIS)

    Information is presented on the following subjects: habitat and vegetation, the avifauna, semi-aquatic and terrestrial vertebrates, and aquatic communities of Steel Creek, species of special concern, and radiocesium in Steel Creek. Two main goals of the study were the compilation of a current inventory of the flora and fauna of the Steel Creek ecosystem and an assessment of the probable impacts of radionuclides, primarily 137Cs, that were released into Steel Creek during earlier reactor operations. Although a thorough evaluation of the impacts of the L reactor restart is impossible at this time, it is concluded that the effects on the Steel Creek ecosystem will be substantial if no mitigative measures are taken

  17. Japan: The institute for the economy of energy recommends a quick re-start of nuclear reactors

    International Nuclear Information System (INIS)

    The Japanese Institute for the Economy of the Energy (IEEJ) considers that the sooner the nuclear reactors will re-start, the better the Japanese economy and environment will be. The 48 Japanese reactors were stopped after the Fukushima accident and their restart is linked to the implementation of new measures for reinforcing safety. Until now only 2 reactors Sendai 1 and Sendai 2 have been allowed to re-start. The procedure for the safety assessment of the reactors is slower than expected. A study shows that only 7 reactors may be allowed to re-start before march 2015 and a total of 19 units may be operating in march 2016. In this scenario 2% of the electricity will come from nuclear energy in 2014 and 15% in 2015, natural gas imports will still be necessary for the production of electricity and their global cost is estimated to reach 56 billions euros while Japan's rate of energy independence will drop by 4.6%. (A.C.)

  18. Mental health effects of the Three Mile Island nuclear reactor restart

    International Nuclear Information System (INIS)

    Controversy over potential mental health effects of the Three Mile Island Unit-1 restart led the authors to examine prospectively the pattern of psychiatric symptoms in a sample of Three Mile Island area mothers of young children. Symptom levels after restart were elevated over previous levels; a sizable subcohort of the sample reported relatively serious degrees of postrestart distress. History of diagnosable major depression and generalized anxiety following the Three Mile Island accident, plus symptoms and beliefs about personal risk prior to the restart, best predicted postrestart symptoms

  19. Cabri - water loop a new IPSN-OECD international research program

    International Nuclear Information System (INIS)

    In 1993, the Institut de Protection et de Surete Nucleaire (IPSN, the French Nuclear Safety and Radiation Protection Institute) working with EDF (electric utilities) and backed by the NRC (Nuclear Regulatory Commission, USA) launched a research program, dubbed Cabri REP Na addressing uranium oxide-based fuels and MOX fuels. So far twelve tests have been conducted including eight on UO2 fuel and four on MOX fuel. More testing is now required to determine fuel performance at higher specific burn-up levels in typical PWR (Pressurized Water Reactor) conditions, the purpose being to determine the acceptance criteria for tomorrow's fuels. IPSN has defined a new research program for the Cabri reactor. The OECD's Nuclear Energy Agency is quarterbacking the international program called 'Cabri-Water Loop'. (authors)

  20. The behaviour of irradiated fuel under RIA transients: Interpretation of the CABRI experiments

    International Nuclear Information System (INIS)

    Paper presents the results of investigation of highly irradiated PWR fuel behaviour under fast power transients conducted in a sodium loop of CABRI reactor, as well as the results on development and validation of computer code SCANAIR. (author). 8 refs, 9 figs, 2 tabs

  1. Core characterization of the new CABRI Water Loop Facility

    International Nuclear Information System (INIS)

    The CABRI experimental reactor is located at the Cadarache nuclear research center, southern France. It is operated by the Atomic Energy Commission (CEA) and devoted to IRSN (Institut de Radioprotection et de Surete Nucleaire) safety programmes. It has been successfully operated during the last 30 years, enlightening the knowledge of FBR and LWR fuel behaviour during Reactivity Insertion Accident (RIA) and Loss Of Coolant Accident (LOCA) transients in the frame of IPSN (Institut de Protection et de Surete Nucleaire) and now IRSN programmes devoted to reactor safety. This operation was interrupted in 2003 to allow for a whole facility renewal programme for the need of the CABRI International Programme (CIP) carried out by IRSN under the OECD umbrella. The principle of operation of the facility is based on the control of 3He, a major gaseous neutron absorber, in the core geometry. The purpose of this paper is to illustrate how several dosimetric devices have been set up to better characterize the core during the upcoming commissioning campaign. It presents the schemes and tools dedicated to core characterization. (authors)

  2. The CABRI facility: Implementation of a pressurized water loop and related safety review

    International Nuclear Information System (INIS)

    The CABRI reactor operated by CEA at the CADARACHE Nuclear Center in France provided the condition for safety studies on nuclear fuel. Initially designed to support investigations on Fast Reactor fuel, large modifications are underway to provide representative conditions for studies on Light Water Fuel types. A general overview of these modifications, the related safety review and supporting studies are described in the paper. (author)

  3. The status of the CABRI test program: Recent results and future activities

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Gonnier, Ch.; Papin, J. [Institut de Protection et de Surete Nucleaire, Saint Paul Lez Durance (France)

    1997-01-01

    The first five CABRI experiments of the REP-Na series, all with UO2 fuel and up to a maximum local burnup of 64 GWd/t, have been examined and analyzed and are now reasonably well understood. In March 1996, the first MOX test with a 3 cycle irradiated fuel at 47 GWd/t radially averaged, local maximum burnup has been successfully performed. The rod did not fail and detailed examinations are being obtained and still in progress presently. The available results and findings are presented in this paper. Three experiments of the REP-Na test matrix are still to be performed, REP-Na7, a 4 cycle MOX test, is scheduled in November 1996. The last two experiments, REP-Na 8 the key experiment of the UO2 matrix, and REP-Na 9, a 2 cycle MOX fuel test, will be performed during the first half of 1997. The CABRI tests made with sodium cooling have a good representativity of reactor conditions during some tens of milliseconds. For better simulation on a longer time range, a project study has been undertaken in view of the implementation of a pressurized-water loop into the CABRI reactor. The design of this loop and the performance parameters of the upgraded driver core of CABRI is presented. Finally, the planning of the CABRI transformation and the outlines of the future test matrix is given. The most optimistic estimation allows to predict that the first tests under prototypical test conditions could be performed before the end of 1999.

  4. Safety of nuclear fuels. New Cabri - water loop IPSN / OECD international research programme

    International Nuclear Information System (INIS)

    A reactivity accident could occur in the case of fuel element accidental ejection. It would induce a quick increase of the neutrons flow in the reactor core, and then to the rupture of pins close to the ejected rod and to the dispersion of the fuel they contain. This phenomenon could lead to give off energy by interaction between the fuel and the primary circuit water. The Cabri research reactor was used to study this case of accident, but this reactor is equipped with a sodium loop; in order to get more representative thermohydraulic conditions for the PWR type reactors, the sodium loop must be replaced by a water loop. Update relative to new radiation protection conditions against the seismic and fire risks is also planned. (N.C.)

  5. Measurement principles of quantitative gamma spectrometry used on Cabri and Phebus testing facilities

    International Nuclear Information System (INIS)

    To improve knowledge about nuclear fuel behavior during accidental sequences and to get information about fission product releases during severe accident in a nuclear reactor, IRSN performs in-pile tests at Cadarache, in the PHEBUS and CABRI experimental facilities. In the frame of these research, IRSN/DRS is steering the CSA(1) research program, which simulates reactivity accident in the CABRI facility, and the PHEBUS FP(2) (fission products) program which simulates in the PHEBUS facility severe accidents on a 20 fuel rods test section and studies fission products release. We developed specific tools to examine the fuel test section before and after a test to get information on the degradation state of the fuel at the end of the transient. The developed instrumentation can measure and localize fission products inside a fuel test section, which is degraded, composed of very dense material and contains strong gamma emitters. Three types of examinations are used concomitantly to provide quantitative results on fission products inside the fuel test section. The first one consists in a gamma scanning along the test device, and provides a first qualitative analysis. The second one is tomography examination which provides information on distribution of high density materials. This technique probes sections of the fuel test device with the use of a high energy X ray source of the LINAC type. After calibration of tomography, 2D images are obtained in which pixel values are converted to densities. Our test devices being very severely degraded after a test, a last examination, called emission computerized tomography, is performed. This technique localizes the different gamma emitters in the test section. This is based on transmission tomography. Gamma emitters are measured but not transmission attenuation. Hyper-pure Germanium collimated mobile detector is used to measure only lines of emission and radial evolutions on multiple angles of measure. About 2000 spectra are

  6. Overview on phenomena of mechanical energy release in the CABRI-experiments

    International Nuclear Information System (INIS)

    Mechanical energy release phenomena observed in the CABRI-experiments are overviewed and discussed. Intensifying and mitigating effects are identified. They are caused by fission gases, inertia effects, pressure dissipation and fissile power

  7. Restarting delayed nuclear power plant projects

    International Nuclear Information System (INIS)

    The management of nuclear power plant projects with delays of several years with respect to the original scheduled commercial operation date presents particular issues and problem areas beyond the normal management tasks needed for projects implemented within originally planned schedules. During the years 1997-1998 the IAEA collected information and practical examples on necessary management actions to preserve the capability for resuming work and completing delayed nuclear power projects when conditions permit. The results were published in the IAEA-TECDOC-1110 entitled Management of Delayed Nuclear Power Plant Projects. As this publication was finalized, at the end of 1998, the available information at the IAEA PRIS (Power Reactor Information System) indicated that more than 40 nuclear power plant projects had delays of five or more years with respect to the originally scheduled operation dates. The PRIS data, collected at the end of 2007, illustrates a trend toward restarting and completing projects that were once delayed. The data indicates that the number of delayed nuclear power projects has reduced to about 25. Some practical methodologies and successful experience from the restarted projects were reviewed, summarized and included in the present publication. The purpose is to address the specific management issues pertaining to a delayed nuclear power project in the period after the decision for restarting is adopted. This publication covers those management issues not considered within the normal processes described in other IAEA publications. It is expected that the practical experience collected from delayed nuclear power projects that were successfully restarted, completed and brought to commercial operation, can provide useful assistance to the management of similar projects considering resumption of work in the future. This publication was produced within the IAEA programme directed to increase the capability of Member States for strengthening national

  8. Colombo cabri or vegetarian meal: wherein lies the power?

    Directory of Open Access Journals (Sweden)

    Meritxell Martín-i-Pardo

    2006-05-01

    Full Text Available “Colombo Cabri or Vegetarian Meal” montre que certains aliments sont utilisés pour définir deux sectes Hindouistes concurrentes en Guadeloupe, dans les Antilles françaises. Ces groupes expriment des identités différentes concurrentes en termes de visibilité et de revendication de pouvoir pour contrôler l’avenir de l’Hindouisme sur l’île. Ces deux entités, identifiées à juste titre comme Hindouistes « traditionalistes » d’une part et « globalistes » d’autre part, mettent en œuvre une rhétorique pour légitimer leurs différentes revendications qui passe par l’adoption ou le rejet du « colombo », un curry de viandes marinées dans cette sauce, comme repas rituel de la secte dont le discours vise à dicter le droit chemin de la communauté Hindouiste de l’île.“Colombo Cabri or Vegetarian Meal” argues that certain foods are used to configure two competing sectarian Hindu groups in Guadeloupe, French West Indies. These groups configure different identities to compete for the attention and claim effective power to control the future of Hinduism on the island. What are appropriately identified as “traditionalist” and “globalist” Hindus define a rhetoric for legitimating their different claims by appropriating or rejecting “colombo,” a curry of meats simmered in this sauce, as the ritual meal for the sect whose narrative rightly claims to define the correct path for the Hindu community on the island.

  9. Utility pushes TMI restart

    International Nuclear Information System (INIS)

    This article discusses management's ability to operate TMI-1 at Three Mile Island nuclear station. The crane that will lift the 170-ton reactor head was the topic of a management safety report, but the Nuclear Regulatory Commission has already approved a series of tests to verify the crane's lifting capacity. Tests consist of lifting four 40-ton concrete missile shields that protect the reactor and one 30-ton concrete shield that protects the pressurizer. They will be placed on a steel load-test frame already on the operating floor

  10. Savannah River Site peer evaluator standards: Operator assessment for restart

    International Nuclear Information System (INIS)

    Savannah River Site has implemented a Peer Evaluator program for the assessment of certified Central Control Room Operators, Central Control Room Supervisors and Shift Technical Engineers prior to restart. This program is modeled after the nuclear Regulatory Commission's (NRC's) Examiner Standard, ES-601, for the requalification of licensed operators in the commercial utility industry. It has been tailored to reflect the unique differences between Savannah River production reactors and commercial power reactors

  11. Restarting TMI unit one: social and psychological impacts

    International Nuclear Information System (INIS)

    A technical background is provided for preparing an environmental assessment of the social and psychological impacts of restarting the undamaged reactor at Three Mile Island (TMI). Its purpose is to define the factors that may cause impacts, to define what those impacts might be, and to make a preliminary assessment of how impacts could be mitigated. It does not attempt to predict or project the magnitude of impacts. Four major research activities were undertaken: a literature review, focus-group discussions, community profiling, and community surveys. As much as possible, impacts of the accident at Unit 2 were differentiated from the possible impacts of restarting Unit 1. It is concluded that restart will generate social conflict in the TMI vicinity which could lead to adverse effects. Furthermore, between 30 and 50 percent of the population possess characteristics which are associated with vulnerability to experiencing negative impacts. Adverse effects, however, can be reduced with a community-based mitigation strategy

  12. Restarting TMI unit one: social and psychological impacts

    Energy Technology Data Exchange (ETDEWEB)

    Sorensen, J.; Soderstrom, J.; Bolin, R.; Copenhaver, E.; Carnes, S.

    1983-12-01

    A technical background is provided for preparing an environmental assessment of the social and psychological impacts of restarting the undamaged reactor at Three Mile Island (TMI). Its purpose is to define the factors that may cause impacts, to define what those impacts might be, and to make a preliminary assessment of how impacts could be mitigated. It does not attempt to predict or project the magnitude of impacts. Four major research activities were undertaken: a literature review, focus-group discussions, community profiling, and community surveys. As much as possible, impacts of the accident at Unit 2 were differentiated from the possible impacts of restarting Unit 1. It is concluded that restart will generate social conflict in the TMI vicinity which could lead to adverse effects. Furthermore, between 30 and 50 percent of the population possess characteristics which are associated with vulnerability to experiencing negative impacts. Adverse effects, however, can be reduced with a community-based mitigation strategy.

  13. Restart Strategies and Internet Congestion

    CERN Document Server

    Maurer, S M; Maurer, Sebastian M.; Huberman, Bernardo A.

    1999-01-01

    We recently presented a methodology for quantitatively reducing the risk and cost of executing electronic transactions in a bursty network environment such as the Internet. In the language of portfolio theory, time to complete a transaction and its variance replace the expected return and risk associated with a security, whereas restart times replace combinations of securities. While such a strategy works well with single users, the question remains as to its usefulness when used by many. By using mean field arguments and agent-based simulations, we determine that a restart strategy remains advantageous even if everybody uses it.

  14. Comments on the possible roles of volatile fission products (cesium) in CABRI tests

    International Nuclear Information System (INIS)

    An investigation of information within the CABRI program that relates to the possible roles of volatile fission products (as represented by cesium) will be described. This study was partially motivated by the observation of localized 137Cs concentration peaks in the axial gamma scans of pins pre-irradiated to about 5% burnup (B.U.) level. In order to evaluate potential effects of such concentrations, a re-examination of the existing test data for the 1% B.U. pins was performed. A comparison of CABRI hodoscope fuel motion results and the pre-CABRI 137Cs axial concentration profiles revealed an approximate spatial correlation between the initial points of fuels dispersal and cesium concentration enhancements (seven of eight cases). 9 figs

  15. Experiences of graduate students: Using Cabri as a visualization tool in math education

    Directory of Open Access Journals (Sweden)

    Çiğdem Gül

    2014-12-01

    Full Text Available Through the use of graphic calculators and dynamic software running on computers and mobile devices, students can learn complex algebraic concepts. The purpose of this study is to investigate the experiences of graduate students using Cabri as a visualization tool in math education. The qualitative case study was used in this study. Five students from graduate students studying at the non-thesis math program of a university located in the Blacksea region were the participant of the study. As a dynamic learning tool, Cabri provided participants an environment where participants visually discovered the geometry. It was concluded that dynamic learning tools like Cabri has a huge potential for teaching visually the challenging concepts that students struggle to image. Further research should investigate the potential plans for integrating the use of dynamic learning software into the math curriculum

  16. Heat exchanger restart evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

    1992-03-18

    On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summary herein.

  17. Heat exchanger restart evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Morrison, J.M.; Hirst, C.W.; Lentz, T.F.

    1992-03-18

    On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized.

  18. Heat exchanger restart evaluation

    International Nuclear Information System (INIS)

    On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4A was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the identified tube. The leaking tube was removed and examined metallurgically to determine the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized

  19. Actividad multisesión con Cabri-Géometre (La circunferencia de Feuerbach)

    OpenAIRE

    Álvarez, José

    1997-01-01

    Se describe una experiencia dirigida al alumnado del segundo ciclo de Enseñanza Secundaria Obligatoria (ESO). Se utiliza el programa didáctico de geometría Cabri Geometre II. Se toma la circunferencia de Feuerbach o circunferencia de nueve puntos como pretexto para la presentación de varios conceptos elementales de geometría plana.

  20. Business Restart in Visegrad Countries

    Directory of Open Access Journals (Sweden)

    Marian HOLIENKA

    2014-03-01

    Our  study  is  the  first  comprehensive  analysis  of  the under-researched topic of business restart in the V4 region. In addition to highlighting its importance as an integral part of entrepreneurial dynamics in V4 countries, it also identifies individual-level drivers of this specifictype of entrepreneurial activity.

  1. LHC Report: Restart preparations continue

    CERN Multimedia

    Katy Foraz for the LHC team and Julia Trummer for the RP Group

    2012-01-01

    Maintenance and consolidation work has been progressing well in both the machine and the experiments in preparation for the March restart.   A sample material is attached to the LHC (the white bag taped to the green line), to measure the radiation doses. Additional work was required around Point 5 due to the discovery and repair of a problem with the RF fingers at the connection of two beam vacuum chambers in CMS. The repair has been completed successfully and the sector is now under vacuum. In order to avoid rushing the delicate final operations required for closing the detector, the restart of the machine has been postponed by one week, from 7 March to 14 March. In the machine, the first cool-down to 1.9 K has started in several sectors ,and the cool-down of the whole machine is still planned to be finished by 21 February. The time window between 22 February and 14 March will be dedicated to powering and cryogenic tests. Since 12 December, the Radiation Protection (RP) group has been deep...

  2. Experiences of graduate students: Using Cabri as a visualization tool in math education

    OpenAIRE

    Çiğdem Gül; Ömür Akdemir; Murat Genç

    2014-01-01

    Through the use of graphic calculators and dynamic software running on computers and mobile devices, students can learn complex algebraic concepts. The purpose of this study is to investigate the experiences of graduate students using Cabri as a visualization tool in math education. The qualitative case study was used in this study. Five students from graduate students studying at the non-thesis math program of a university located in the Blacksea region were the participant of the study. As ...

  3. Public communication toward Monju restart

    International Nuclear Information System (INIS)

    Five years have gone by since the sodium leak took place at a prototype FBR Monju. Looking back upon that time, one journalist said, The Monju accident was technically far from the serious one as being reported in the media. Had it not been for the infamous 'accident cover-up', an uproar must have calmed down in a month. But an unexpectedly large negative public reaction has kept Monju idle all these years. What had really happened? There was a false report on the time of first entry to the piping room or the sodium-leak spot. Contrary to the fact that five staffs did enter the room at 2:00 am, PNC failed to mention it at a first press conference held at 8:30 am. Instead, PNC created a fictitious time of entry at 10:00 am and reported it to the authorities in a formal document. Another mishap was a video cover-up operation. A year and three months later, an explosion accident took place at PNC's Tokai Reprocessing Facility and similar mishap was repeated then, causing a fatal damage to the PNC's reputation. Public opinion polls taken by mass media have concluded that PNC is 'bureaucratic, closed, slow in coping with situation and untrustworthy'. PNC struggle began - struggle to regain public trust. A series of mishaps at PNC have created an anxiety and distrust about nuclear energy among the nation. In order to restore the trust of the nation, STA, a government agency supervising PNC, decided that PNC be reorganised to make a new start as Japan Nuclear Cycle Development Institute (JNC) on October 1, 1998. In the start of the new organisation, JNC is expected to carry out operations placing priority on the locality of its facilities. The most precious lesson learned from the Monju accident is the importance of public communication. Currently undertaking activities toward Monju restart are; 1) public opinion monitoring, 2) social meetings, 3) strengthening publicity activities, 4) dialogue with local administration officials and opinion leaders, 5) 'open meeting

  4. Preliminary study on restarts for CSP algorithms

    OpenAIRE

    Baptista, Luís

    2010-01-01

    O uso de técnicas de restarts para resolver problemas de satisfação de restrições (CSPs), utilizando algoritmos de procura com retrocesso, é considerado pouco importante. Neste artigo propomos conduzir um estudo preliminar sobre o impacto da utilização de restarts nestes algoritmos. Mostramos que o conhecido problema da n-rainhas tem uma distribuição heavy-tail. Apresentamos evidências empíricas de que os restarts podem efectivamente melhorar o tempo necessário para encontrar a...

  5. LHC Report: A tough restart

    CERN Multimedia

    Jan Uythoven for the LHC team

    2012-01-01

    The third LHC Technical Stop of five days took place in the week of September 17. Getting back to normal operation after a technical stop  can sometimes be difficult, with debugging, testing and requalification required on the systems that have seen interventions. Folding in a selection of other problems can make for a frustrating time.   The new injector magnet is transported to the LHC. Photo: TE/ABT group. The restart experienced over the last days was one of the tougher ones. Many problems occurred, both small and large, one after the other; in the end it took until Sunday afternoon, 9 days after the end of the technical stop, to have a physics fill in the machine that delivered an initial luminosity similar to those before the technical stop. Most problems encountered were, in fact, not related to the technical stop. The technical stop consisted of the usual maintenance and consolidation of the various systems, but two items stand out: the replacement of the mirrors an...

  6. La geometría dinámica con CABRI II

    OpenAIRE

    Barrales, Marco; Carral, Michel

    2004-01-01

    La utilización de una herramienta nueva, de cualquier tipo que sea, necesita de una reflexión sobre lo que hacemos, muchas veces cambia nuestro modo de trabajar (actitud) y hace surgir problemas sobre las verdades que teníamos. En matemática los conocimientos utilizados pueden ser diferentes: comparar una construcción geométrica con regla y compás o con regla y escuadra (mecánica) o solamente con compás. En este curso se explora de manera activa el software Cabri II. En una primera etapa se r...

  7. Entre el plano y el espacio la dimensión fractal con Cabri

    OpenAIRE

    Cánepa, Ana Inés; Fayó, Alicia Noemí

    2007-01-01

    Se trabajará mediante el método Aula – Taller con guías de trabajos prácticos que inducirán a los docentes a investigar en Cabri los temas a desarrollar. El taller está dirigido para docentes de nivel medio, terciario que deseen incorporar el relevante tema de Fractales en la curricula Mediante la observación de un video sobre Fractales y lectura de textos sobre el tema se invitará a los asistentes a recorren este nuevo mundo que permite desde la simplicidad de un elemento geométrico llegar a...

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

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  9. Restart of protype FBR 'Monju' after long-term shut-down

    International Nuclear Information System (INIS)

    The sodium leak accident in a secondary main cooling system of prototype fast breeder reactor Monju on December 8th in 1995, and the Monju has shut down for 14 and half years since that JAEA improved the safety by investigating the sodium leakage accident, taking countermeasures such as better understanding by the local's, improvement of operation. In the maintenance field, the confirmation of Monju systems and components integrity, systems improvement as sodium leak countermeasures, introduction of the maintenance program to resolve maintenance problems. Monju restarted on May 6th 2010 after the confirmation of restart readiness by safety authorities and the core confirmation test which is first test after restart was continued to July 22nd as planned. (author)

  10. Risk perception in an interest group context: an examination of the TMI restart issue

    International Nuclear Information System (INIS)

    Human response to environmental hazards and risks has been the subject of considerable research by social scientists. Work has traditionally focused on either individual response to the risks of an ongoing or future threat (hazards research), or group and organizational response to a specific disaster event (disaster research). As part of a larger investigation of the restart of the Unit 1 reactor at Three Mile Island (TMI), the response of interest groups active in the restart issue to the continued threat of TMI and to future risks due to restart was examined. After reviewing the restart issue in general, the local dimensions of the restart issue from interest group perspectives are discussed. A method for defining appropriate issues at the community level is reviewed. Differences in the perceived local impacts of alternative decisions, and systems of beliefs associated with differing perceptions are discussed. Finally, the implications of interest group versus individual perceptions of local issues for decision making about TMI, in particular, and about technological hazards management, in general, are discussed. Associated implications for determining socially acceptable risk levels are identified

  11. Methods used and kind of results obtained in SCARABEE facility about fuel and clad motion diagnostics. Previsional development for SCARABEE N and CABRI facilities

    International Nuclear Information System (INIS)

    Techniques are described for planned LMFBR loss of flow simulation studies using irradiated fuel (SCARABEE N program) and for LMFBR transient overpower studies using both fresh and irradiated fuel (CABRI program)

  12. Beleaguered LHC gears up for restart

    CERN Multimedia

    Cartwright, Jon

    2009-01-01

    "The Large Hadron Collider (LHC) is finally set to restart in mid-November following last year's accident. Initially it will collide protons at an energy of only 3.5 TeV per beam, and staff at Cern will have to wait until late next year before trying to run the collider at its maximum energy" (0.75 page)

  13. LHC Experiments: refinements for the restart

    CERN Multimedia

    2009-01-01

    As the LHC restart draws closer, the Bulletin will be taking a look at how the six LHC experiments are preparing and what they have been up to since last September. In this issue we start with a roundup of the past 10 months of activity at CMS and ATLAS, both technical work and outreach activities.

  14. Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

    International Nuclear Information System (INIS)

    Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

  15. Restart of the LHC in 2009

    CERN Multimedia

    Corinne Pralavorio

    The restart of the LHC during the summer 2009 has been confirmed today, the 5 December. An updated report on the incident which damaged sector 3-4 has just been published. It gives details on the damage caused by the incident and explains the ongoing repairs and the new systems being put into place to reinforce the safety of the machine. Click here to see the report.

  16. Restartable solid motor stage for shuttle applications

    Science.gov (United States)

    Rohrbaugh, D. J.

    1973-01-01

    The application of restartable solid motor stages to shuttle missions has been shown to provide a viable supplement to the shuttle program. Restartable solid motors in the 3000 pound class provide a small expendable transfer stage that reduces the demand on the shuttle for the lower energy missions. Shuttle operational requirements and preliminary performance data provided an input for defining design features required for restartable solid motor applications. These data provided a basis for a configuration definition that is compatible with shuttle operations. Mission by mission analysis showed the impact on a NASA supplied mission model. The results showed a 15% reduction in the number of shuttle flights required. In addition the amount of shuttle capability used to complete the mission objectives was significantly reduced. For example, in the 1979 missions there was a 62% reduction in shuttle capability used. The study also showed that the solid motor could provide a supplement to the TUG that would allow TUGS to be used in a recoverable rather than an expendable mode. The study shows a 71% reduction in the number of TUGs that would be expended.

  17. Waxy crude oil flow restart ability

    Energy Technology Data Exchange (ETDEWEB)

    Sierra, Andre Gaona; Varges, Priscilla Ribeiro; Mendes, Paulo Roberto de Souza [Dept. of Mechanical Engineering. Pontificia Universidade Catolica do Rio de Janeiro, RJ (Brazil)], e-mails: prvarges@puc-rio.br, pmendes@puc-rio.br; Ziglio, Claudio [PETROBRAS S.A, R.J., Rio de Janeiro, RJ (Brazil)], e-mail: ziglio@petrobras.com.br

    2010-07-01

    Under the hot reservoir conditions, waxy crudes behave like Newtonian fluids but once they experience very cold temperatures on the sea floor, the heavy paraffin's begin to precipitate from the solution impacting non- Newtonian flow behavior to the crude (Chang 2000, Lee 2009, Davidson 2004) and begin to deposit on the pipe wall leave blocked of pipeline. This gel cannot be broken with the original steady state flow operating pressure applied before gelation (Chang 1998). Restarting waxy crude oil flows in pipelines is a difficult issue because of the complex rheological behavior of the gelled oil. Indeed, below the WAT, the gelled oil exhibits viscoplastic, thixotropic, temperature-dependent, and compressible properties due to the interlocking gel-like structure formed by the crystallized paraffin compounds and the thermal shrinkage of the oil. The main objective of this work is to determine the minimal pressure to restart the flow, and the relationship between the fluid rheology , pipe geometry and the restart pressure of the flow. Experiments will be performed to investigate the displacement of carbopol aqueous solutions (viscoplastic fluid without thixotropic effects) by Newtonian oil flowing through a strait pipe to validate the experimental apparatus. Therefore, tests will be made with different fluids, like Laponite and waxy crude oils. (author)

  18. Estudio de las funciones trigonométricas con Cabri: una estrategia para su enseñanza

    OpenAIRE

    Betín, Erika; Gonzales, Ivan; Oviedo, Elver

    2004-01-01

    A través del taller se muestra la posibilidad del uso del programa computacional Cabri para el desarrollo del pensamiento variacional especialmente; mostrando el comportamiento general de cada una de las funciones trigonométricas en el plano cartesiano, graficándolas en el mismo plano haciendo una simulación de eje “y” sobre el mismo sistema coordenado.

  19. 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)

  20. The project to implement into CABRI a pressurized water loop. Motivations and objective of the future test program

    International Nuclear Information System (INIS)

    Recent high burnup fuel tests in NSRR and in CABRI have invalidated the former fuel design acceptance criteria. In particular, there is no longer any doubt that with increasing burnup, the transient mean fuel enthalpy failure and dispersal of the fuel decreases significantly. The physical phenomena must be understood and modelled, post-failure phenomena must be evaluated under the conditions of a representative environment. Three major aspects are discussed: identification of arguments which show the need of a pressurised water-loop in CABRI; review of the performance and diagnostic requirements in the perspective of a future RIA test program; presenting the outcome of the technological feasibility study. A pressurised water environment together with representative flow and temperature conditions is needed, especially for the study of post-DNB and post-failure behaviour of the highly irradiated LWR fuel. A detailed study has demonstrated the feasibility of the implantation of a pressurised water loop into CABRI; the safe operation of the experimental facility and the achievement of the technical requirements will be possible. (R.P.)

  1. Response for flooding and effort of plant restart at Fort Calhoun NPP

    International Nuclear Information System (INIS)

    Fort Calhoun NPP(FCNPP) experienced flood due to the elevation of water level of Missouri river from June to August, 2011. FCNPP took various responses including constructing the berm around the important facilities, and could maintain the safety of the reactor with difficulty. The response of FCNPP, which includes both good and bad aspects, will help us in examining the response to flooding at Japanese nuclear plants. In response to the occurrence of the fire of safety related breaker at FCNPP, June 2011, NRC entered the oversight for FCNPP based on the Inspection Manual Chapter-0350 and issued the Confirmatory Action Letter (CAL) with the check list to Omaha Public Power District (OPPD), owner of FCNPP. Check list described the items which OPPD should solve before the restart of FCNPP. Although many efforts need to be done by OPPD to solve the check list, there is transparency about the condition of the plant restart. In Japan all the plants except Ohi 3,4 have continued to shut down after Fukushima accident due to the lack of the rule for restarting of the plant after the occurrence of the safety problem. We should study the plant restart system of NRC. (author)

  2. 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

  3. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-12-01

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

  4. LHC Report: Rocky re-start

    CERN Multimedia

    Barbara Holzer for the LHC Team

    2012-01-01

    A rocky re-start with beam followed a successful machine development period and the first technical stop of 2012. Today, Friday 11 May, the machine began running again with 1380 bunches.   A short, two-day machine development period was successfully completed on 21-22 April. It focused on topics relevant for the 2012 physics beam operation. This was then followed by a five-day technical stop, the first of the year. The technical stop finished on time on Friday 26 April. The re-start with beam was somewhat tortuous and hampered by an unlucky succession of technical faults leading to extended periods of downtime. The planned intensity increase was put on hold for three days with the machine operating with 1092 bunches and a moderate bunch intensity of 1.3x1011 protons. This delivered a reasonable peak luminosity of 3.6x1033 cm-2s-1 to ATLAS and CMS. Higher than usual beam losses were observed in the ramp and squeeze, and time was required to investigate the causes and to implement mitigati...

  5. Numerical simulations of a restarting jet

    International Nuclear Information System (INIS)

    The first self-consistent two-dimensional numerical simulation of an axisymmetric intermittent jet is presented. It is found that if the jet parameters at the jet nozzle do not vary widely from one duty cycle to the next, the restarted jet will always be overdense if the original jet is underdense relative to the quiescent intergalactic medium. Although the working surface advance speed of the restarted jet is greater than that of the original jet, the Mach number of the advance speed is less. Because the advance speed Mach number of the new working surface is relatively low, the bow shock excited by the new jet is weak regardless of the strength of the original bow shock in the quiescent ambient medium. Because the bow shock and terminal Mach disk must have comparable strengths, the brightness contrast of emission features associated with the shock should be roughly the same. Without the momentum flux of the jet, the hot spot expands in a time short compared with the duty cycle. 27 refs

  6. Cold restart of viscous multiphase flowline by hot water flushing

    NARCIS (Netherlands)

    Yang, Z.; Velthuis, J.F.M.; Veltin, J.; Twerda, A.

    2013-01-01

    The main objective of this paper is to understand the physical processes of shut-in and restart of viscous multiphase flowline. For this purpose, experiments were performed on shut-in and restart processes in the Multiphase Flow Rig at Statoil's research Centre in Porsgrunn. Real crudes were used wi

  7. 40 CFR 86.536-78 - Engine starting and restarting.

    Science.gov (United States)

    2010-07-01

    ... throttle about half way and cranking the engine until it starts. (b) (c) If, during the cold start, the... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Engine starting and restarting. 86.536... Regulations for 1978 and Later New Motorcycles; Test Procedures § 86.536-78 Engine starting and restarting....

  8. 40 CFR 86.136-90 - Engine starting and restarting.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Engine starting and restarting. 86.136... Complete Heavy-Duty Vehicles; Test Procedures § 86.136-90 Engine starting and restarting. (a) Otto-cycle vehicles. Paragraph (a) of this section applies to Otto-cycle vehicles. (1) The engine shall be...

  9. 40 CFR 1065.930 - Engine starting, restarting, and shutdown.

    Science.gov (United States)

    2010-07-01

    ... otherwise, start, restart, and shut down the test engine for field testing as follows: (a) Start or restart the engine as described in the owners manual. (b) If the engine does not start after 15 seconds of cranking, stop cranking and determine the reason it failed to start. However, you may crank the...

  10. Cabri II Plus como herramienta para la enseñanza de las isometrías

    OpenAIRE

    Díaz-Levicoy, Danilo; Juan C Sánchez; Mayorga, Alexis

    2013-01-01

    La incorporación en la vida cotidiana de las nuevas tecnologías de la información y la comunicación ha significado un cambio radical en la forma de desarrollar el proceso de enseñanza y aprendizaje en las diferentes disciplinas y niveles escolares. En este sentido, el software de geometría dinámica “Cabri Géomètre II Plus” es un programa computacional de fácil manipulación, amigable y de rápido aprendizaje, que permite a los estudiantes visualizar, descubrir, conjeturar y/o comprobar propieda...

  11. Re-starting an Arnoldi iteration

    Energy Technology Data Exchange (ETDEWEB)

    Lehoucq, R.B. [Argonne National Lab., IL (United States)

    1996-12-31

    The Arnoldi iteration is an efficient procedure for approximating a subset of the eigensystem of a large sparse n x n matrix A. The iteration produces a partial orthogonal reduction of A into an upper Hessenberg matrix H{sub m} of order m. The eigenvalues of this small matrix H{sub m} are used to approximate a subset of the eigenvalues of the large matrix A. The eigenvalues of H{sub m} improve as estimates to those of A as m increases. Unfortunately, so does the cost and storage of the reduction. The idea of re-starting the Arnoldi iteration is motivated by the prohibitive cost associated with building a large factorization.

  12. ALICE & LHCb: refinements for the restart

    CERN Multimedia

    2009-01-01

    Following the previous issue, the Bulletin continues its series to find out what the six LHC experiments have been up to since last September, and how they are preparing for the restart. Previously we looked at CMS and ATLAS; this issue we will round up the past 10 months of activity at ALICE and LHCb. LHCb The cavern of the LHCb experiment. This year has given LHCb the chance to install the 5th and final plane of muon chambers, which will improve the triggering at nominal luminosity. This is the final piece of the experiment to be installed. "Now the detector looks exactly as it does in the technical design report," confirms Andrei Golutvin, LHCb Spokesperson. "We also took advantage of this shutdown to make several improvements. For example, we modified the high voltage system of the electromagnetic calorimeter to reduce noise further to a negligible level. We also took some measures to improve ...

  13. Design criteria and description of El Cabri l I and LLRW repository

    International Nuclear Information System (INIS)

    The facilities at El Cabri l are designed as a low and intermediate level waste near surface disposal facility. They must meet two basic objectives: - Ensure the immediate and deferred protection of the public and the environment - Allow free use of the site after a maximum of 300 year, without any radiological limitation. - Facilitate the retrievability of the wastes if circumstances were to make it advisable. The Disposal system The waste packages, most of them are drums, are place inside concrete disposal containers, where are immobilised with mortar forming a concrete block. These containers are stored in disposal cells. The containers are place in contact with each other, a central cross being left to allow for container manufacturing or positioning tolerances. Once each disposal cell has been fully loaded the central strip is backfilled with gravel to fill the gaps, and upper closing slab is built. The bottom plate of the cell collects any seepage water and channels it to a network of pipes installed in inspection drifts located below the disposal cells. Each cell is linked to this network, called Infiltration Control Network. via a holding tank, so that if water is collected, it is possible to know which disposal cell it has come from, in order to repair the protective covering, and to take samples of the water collected. During the operating phase, and with the triple objective of protecting the waste containers from the weather, minimizing the amount of water collected in the Infiltration Control Network and acting in support of the container hang ling system, each row of disposal cell is served by a rail-mounted sliding roof. When all the cells will be closed down, they will be topped with a low permeability cap, formed by alternating layers of waterproof and draining material, which protect the waste packages in the long term

  14. Shutdown channels and fitted interlocks in atomic reactors

    International Nuclear Information System (INIS)

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors)

  15. Optimal Stochastic Restart Renders Fluctuations in First Passage Times Universal

    Science.gov (United States)

    Reuveni, Shlomi

    2016-04-01

    Stochastic restart may drastically reduce the expected run time of a computer algorithm, expedite the completion of a complex search process, or increase the turnover rate of an enzymatic reaction. These diverse first-passage-time (FPT) processes seem to have very little in common but it is actually quite the other way around. Here we show that the relative standard deviation associated with the FPT of an optimally restarted process, i.e., one that is restarted at a constant (nonzero) rate which brings the mean FPT to a minimum, is always unity. We interpret, further generalize, and discuss this finding and the implications arising from it.

  16. Restart and progress of system start-up test in MONJU

    International Nuclear Information System (INIS)

    The Japanese prototype fast breeder reactor Monju restarted its system startup test(SST) in May 2010 after a 14-year interruption. The SST is planned to be conducted in the three steps: Core Confirmation Test (CCT), 40%-output Confirmation Test, and Power Rising Test. After finishing these steps, the full power operation will start. CCT, which was composed of 20 test items including confirmation of the safety criteria and measurement of several core performance data, finished on July 22nd, 2010 after 78 days test. Through CCT, were demonstrated the stable operation and the prediction accuracy on the core performance after long-term shutdown and refueling. The loaded fuel of Monju is mixed plutonium-uranium oxide with fissile plutonium enrichment around 15-21%. The accumulation of 241Am due to the 241Pu decay during the interruption reaches 1.5wt% in average. An impact of the reactor physics data obtained in the restart core is investigated by the cross section adjustment technique with JENDL-3.3 and JENDL-4.0. Criticality data obtained before and after the interruption are applied. It is confirmed that Monju reactor physics data, when the two data are used together, effectively adjust 241Am capture cross sections. (author)

  17. Press Conference: LHC Restart, Season 2

    CERN Document Server

    CERN. Geneva

    2015-01-01

    PRESS BRIEFING ON THE LARGE HADRON COLLIDER (LHC) RE-START, SEASON 2 AT CERN, GLOBE OF SCIENCE AND INNOVATION Where :   http://cern.ch/directions   at the Globe of Science and Innovation When : Thursday, 12 March from 2.30 to 3.30pm - Open seating as from 2.15pm Speakers : CERN’s Director General, Rolf Heuer and Director of Accelerators, Frédérick Bordry, and representatives of the LHC experiments Webcast : https://webcast.web.cern.ch/webcast/ Dear Journalists, CERN is pleased to invite you to the above press briefing which will take place on Thursday 12 March, in the Globe of Science and Innovation, 1st floor, from 2.30 to 3.30pm. The Large Hadron Collider (LHC) is ready to start up for its second three-year run. The 27km LHC is the largest and most powerful particle accelerator in the world operating at a temperature of -217 degrees Centigrade and powered to a current of 11,000 amps. Run 2 of the LHC follows a two-year technical s...

  18. Current status of JMTR for restart

    International Nuclear Information System (INIS)

    After the 2011 off the Pacific Coast of Tohoku Earthquake on March 11, 2011, JMTR has been challenging to the inspection of facility equipment, seismic soundness evaluation through earthquake response analysis, repair of facilities, correspondence to the report matters stipulated by laws and regulations, and improvement of irradiation facilities, and based on these, it have formulated the operation schedule from FY2013. In the future, JMTR will explain the soundness of the facilities to the Nuclear Regulatory Commission, and receive the facility's regular inspection after the completion of piping updating work related to the report matters stipulated by laws and regulations. After obtaining the understanding of local municipalities, it aims to restart the facilities in August 2013. After the reoperation, it will make efforts to safely and stably operate the facilities with a target of the operating rate of world top class. In addition, the following challenges are planned: (1) improvement of the facilities based on external funds for expanding utilization, (2) human resource development in the nuclear field by utilizing JMTR, and (3) development of an advanced monitoring system for improving the safety of nuclear power plants by utilizing the irradiation technology that has been accumulated in JMTR. With the aim for JMTR becoming the international hub, JMTR will continue aggressive activities. (A.O.)

  19. Preparing for the re-start

    CERN Multimedia

    2009-01-01

    The end of a Council week is a good opportunity to bring you up to date with the status of the LHC, and I’m pleased to say that we had a good deal of positive news to report to the delegations today. The bottom line is that we remain on course to restart the LHC safely this year, albeit currently about 2-3 weeks later than we’d hoped at Chamonix. This Council week has seen many important developments for our future. I am particularly pleased that Council approved the Medium Term Plan and budget for 2010 as presented by the management. This is a strong vote of confidence in all of you. The President of Council is reporting on Council business in this issue of the Bulletin, so I will focus on the status of the LHC. A tremendous amount of work has been done to understand fully the splices in the LHC’s superconducting cable and copper stabilizers. One of these splices was the root cause of the incident last September that brought ...

  20. TOTEM and LHCf: refinements for the restart

    CERN Multimedia

    2009-01-01

    Following the previous two issues, the Bulletin continues its series to find out what the six LHC experiments have been up to since last September, and how they are preparing for the restart. We covered CMS, ATLAS, LHCb and ALICE in previous issues. In this issue we will round up the past 10 months of activity at TOTEM and LHCf. Roman Pots of the TOTEM experiment.TOTEM The past 10 months at TOTEM have been amongst the busiest since the project’s inception. The delay in the LHC startup has certainly had a silver lining for the TOTEM collaboration - not only has it given them a much-needed opportunity to test and install many crucial new detector parts, but also the lower energy range that the LHC will initially operate at in 2009 is perfect for TOTEM physics. "In fact, the LHC almost seems to be following the schedule of TOTEM!" jokes Karsten Eggert, TOTEM spokesperson. TOTEM is made up of three different detectors spread out...

  1. 40 CFR 86.1236-85 - Engine starting and restarting.

    Science.gov (United States)

    2010-07-01

    ...) Evaporative Emission Test Procedures for New Gasoline-Fueled, Natural Gas-Fueled, Liquefied Petroleum Gas-Fueled and Methanol-Fueled Heavy-Duty Vehicles § 86.1236-85 Engine starting and restarting. (a)...

  2. LHC Availability 2016: Restart to Technical Stop 1

    CERN Document Server

    Todd, Benjamin; Apollonio, Andrea; CERN. Geneva. ATS Department

    2016-01-01

    This document summarises the LHC machine availability for the period of Restart to Technical Stop 1 (TS1) in 2016. This period was dedicated to restart of the LHC for proton physics with a bunch spacing of 25ns. This note has been produced and ratified by the LHC Availability Working Group which has complied fault information for the period in question using the Accelerator Fault Tracker.

  3. Dissociating restart cost and mixing cost in task switching

    OpenAIRE

    Poljac, E; Koch, I.; H. Bekkering

    2009-01-01

    Three experiments investigated the cognitive mechanisms underlying the restart cost and mixing cost in task switching. To this aim, the predictability of task order was varied (unpredictable in Experiment 1 and predictable in Experiments 2 and 3) across experiments, which employed a multiple-trial paradigm. Verbal cues for color and shape matching tasks were presented before a run of four trials. Focusing on task-repetition runs only, we measured restart cost as the difference in performance ...

  4. Checkpointing for the RESTART problem in Markov networks

    OpenAIRE

    Lipsky, Lester; Doran, Derek; Gokhale, Swapna

    2011-01-01

    We apply the known formulae of the RESTART problem to Markov models of software (and many other) systems, and derive new equations. We show how checkpoints might be included, with their resultant performance under RESTART. The result is a complete procedure for finding the mean, variance, and tail behavior of the job completion time as a function of the failure rate. We also provide a detailed example.

  5. Thick-Restart Lanczos Method for Electronic Structure Calculations

    International Nuclear Information System (INIS)

    This paper describes two recent innovations related to the classic Lanczos method for eigenvalue problems, namely the thick-restart technique and dynamic restarting schemes. Combining these two new techniques we are able to implement an efficient eigenvalue problem solver. This paper will demonstrate its effectiveness on one particular class of problems for which this method is well suited: linear eigenvalue problems generated from non-self-consistent electronic structure calculations

  6. A simple strategy for varying the restart parameter in GMRES(m)

    Energy Technology Data Exchange (ETDEWEB)

    Baker, A H; Jessup, E R; Kolev, T V

    2007-10-02

    When solving a system of linear equations with the restarted GMRES method, a fixed restart parameter is typically chosen. We present numerical experiments that demonstrate the beneficial effects of changing the value of the restart parameter in each restart cycle on the total time to solution. We propose a simple strategy for varying the restart parameter and provide some heuristic explanations for its effectiveness based on analysis of the symmetric case.

  7. DÉMARCHE EXPÉRIMENTALE, VALIDATION, ET OSTENSIFS INFORMATISÉS. IMPLICATIONS DANS LA FORMATION D'ENSEIGNANTS À L'UTILISATION DE CABRI EN CLASSE DE GÉOMÉTRIE.

    OpenAIRE

    Acosta, Martin

    2008-01-01

    WE STUDY THE DIFFICULTIES THAT TEACHERS EXPERIENCE IN LEARNING TO USE CABRI IN THEIR TEACHING, AS SEVERAL STUDIES HAVE REVEALED, IN ORDER TO IMPROVE TEACHER'S TRAINING. WE USE A TRIPLE THEORETICAL FRAME, WITH THE ANTHROPOLOGICAL THEORY OF DIDACTICS (CHEVALLARD), THE INSTRUMENTAL APPROACH (RABARDEL) AND THE THEORY OF SITUATIONS (BROUSSEAU). WE HAVE IMPLEMENTED A LONG INGENIERY, WITH AN INTENSIVE COURSE ABOUT GEOMETRIC PROBLEM SOLVING WITH CABRI, A PERIOD OF COACHED PRACTICE, AN INTENSIVE COURS...

  8. Reactor BR2

    International Nuclear Information System (INIS)

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

  9. Bruce A restart (execution and lessons-learned)

    International Nuclear Information System (INIS)

    Lessons learned with the Bruce Units 3 and 4 restart have been incorporated into the current refurbishment of Units 1 and 2. In addition, lessons learned on the lead unit (U2) are aggressively applied on the lagging unit (U1) to maximize efficiency and productivity. There will be a discussion on how this internal OPEX, along with external lessons learned, are used to continuously improve all aspects of the Bruce A Restart project management cycle, from scope selection, through planning and scheduling, to execution.

  10. Restart plan for the prototype vertical denitration calciner

    Energy Technology Data Exchange (ETDEWEB)

    SUTTER, C.S.

    1999-09-01

    Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

  11. 40 CFR 86.236-94 - Engine starting and restarting.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Engine starting and restarting. 86.236-94 Section 86.236-94 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... New Medium-Duty Passenger Vehicles; Cold Temperature Test Procedures § 86.236-94 Engine starting...

  12. Dissociating restart cost and mixing cost in task switching

    NARCIS (Netherlands)

    Poljac, E.; Koch, I.; Bekkering, H.

    2009-01-01

    Three experiments investigated the cognitive mechanisms underlying the restart cost and mixing cost in task switching. To this aim, the predictability of task order was varied (unpredictable in Experiment 1 and predictable in Experiments 2 and 3) across experiments, which employed a multiple-trial p

  13. Restart plan for the prototype vertical denitration calciner

    International Nuclear Information System (INIS)

    Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing

  14. Investigation of the Development of 7th Grade Students’ Skills to Define, Construct and Classify Polygons with Cabri Geometry

    Directory of Open Access Journals (Sweden)

    Ahmet Yanık

    2013-03-01

    Full Text Available The aim of the study is to investigate the development of 7th Grade students’ skills to define, construct and classify polygons in geometry course with Cabri Geometry II Plus software geometry, an example of dynamic geometry software. The study used qualitative and quantitative research methods in accordance with the research objectives and focus, so it was designed as a mixed method research. The participants of the study were 21 7th Grade students, 11 girls and 10 boys, who were attending a secondary school in Eskişehir city center during 2012-2013 school year. As a source of qualitative data, four students in this class were selected for the interview. The data were collected with “Polygon Identification and Classification Scale”, one group pre-test and post-test in order to determine the level of development and significance level of the gender variable, and Cabri Geometry worksheets developed by the researchers. The quantitative data were analyzed with SPSS Statistics 20. Also, t-test and Wilcoxon test were used in data analysis. The data obtained from the interviews were analyzed through descriptive analysis. The qualitative data showed that the mean of correct answers given by the students to the questions in the Polygon Identification and Classification Scale was higher in the post-test than the pre-test. The ttest results for the pre-test and post-test mean scores and the results of the paired samples test showed a significant difference in favor of the post-test. There was no significant difference based on the gender variable. On the other hand, the data obtained from the interviews were coded under five different themes. The activities about the concept of formation showed that incorrect formations caused incorrect generalizations about the shapes. The study found that, as a result of the teaching practice in the study, hierarchical relations among polygons were expressed correctly. Finally, after the practice, the

  15. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors)

  16. Training report of the FBR cycle training facility for Monju Restarting

    International Nuclear Information System (INIS)

    The FBR Cycle Training Facility consists of the sodium handling training facility and the maintenance training facility has been used since September in 2000. At the sodium handling training facility, the trainees are able to study widely sodium handling technologies, which are inherent technologies of fast reactors, such as sodium fire extinguishing, chemical and physical properties of sodium, sodium loop operating skill, counter training for sodium piping leakage, etc.. On the other hand, maintenance training courses not only for 'MONJU' related inherent maintenance technologies but also for general maintenance skills are conducted at the maintenance training facility. So far, 77 eight-unit sodium training courses and 39 nine-unit maintenance training courses have been performed in preparation of Monju restarting and the total number of trainees is 888. (author)

  17. MUS81-EME2 Promotes Replication Fork Restart

    Directory of Open Access Journals (Sweden)

    Alessandra Pepe

    2014-05-01

    Full Text Available Replication forks frequently stall at regions of the genome that are difficult to replicate or contain lesions that cause replication blockage. An important mechanism for the restart of a stalled fork involves endonucleolytic cleavage that can lead to fork restoration and replication progression. Here, we show that the structure-selective endonuclease MUS81-EME2 is responsible for fork cleavage and restart in human cells. The MUS81-EME2 protein, whose actions are restricted to S phase, is also responsible for telomere maintenance in telomerase-negative ALT (Alternative Lengthening of Telomeres cells. In contrast, the G2/M functions of MUS81, such as the cleavage of recombination intermediates and fragile site expression, are promoted by MUS81-EME1. These results define distinct and temporal roles for MUS81-EME1 and MUS81-EME2 in the maintenance of genome stability.

  18. Hoechst Celanese picks Methanex as partner for methanol restart

    Energy Technology Data Exchange (ETDEWEB)

    Morris, G.D.L.

    1992-08-19

    Hoechst Celanese Chemical (Dallas) has chosen Methanex (Vancouver)-the former Ocelot Industries-as its partner for the restart of its long-mothballed 600,000-m.t./year methanol unit at Clear Lake, TX. The two firms are in the final stages of negotiations, and a formal announcement on the agreement could come as early as this week. Interest in methanol, has strengthened dramatically in the past couple of years, reflecting strong demand to feed production of fuel oxygenates.

  19. Applications of implicit restarting in optimization and control Dan Sorensen

    Energy Technology Data Exchange (ETDEWEB)

    Sorensen, D. [Rice Univ., Houston, TX (United States)

    1996-12-31

    Implicit restarting is a technique for combining the implicitly shifted QR mechanism with a k-step Arnoldi or Lanczos factorization to obtain a truncated form of the implicitly shifted QR-iteration suitable for large scale eigenvalue problems. The software package ARPACK based upon this technique has been successfully used to solve large scale symmetric and nonsymmetric (generalized) eigenvalue problems arising from a variety of applications.

  20. The dynamic storage and restart facilities in MABEL-2

    International Nuclear Information System (INIS)

    MABEL-2 is a FORTRAN program for calculating clad ballooning in a PWR during a LOCA. Originally written with fixed array storage, the use of the code has been extended by including dynamic storage. The lengths of the arrays in the program are set at execution time, varying from run to run. This allows much greater freedom in the choice of mesh and the size of case run. The use of computer memory is also more efficient. In addition a restart facility has been included which allows the user to break off and restart execution of the program (once or many times) during a transient. By using this facility much longer calculations can be run. Should an error in either input data or program become apparent late in a transient, the case need only be re-run from the last dump because some input data can be altered at restart. The use of these new facilities and the coding changes are described. (author)

  1. Interpretation of the CABRI-RAFT RB1 and RB2 tests through detailed date evaluation and PAPAS-2S code analysis

    OpenAIRE

    深野 義隆; 佐藤 一憲

    2001-01-01

    The CABRI-RAFT RB1 and RB2 tests were aiming at a study on impact of fuel pin failure under an overpower condition leading to fuel melting. Using a special technique, combination of through-cladding failure and fuel melting was realized.In the RB1 rest, fuel ejection was prevented under a limited fuel melting condition. On the other hand, significant fuel melting was applied in the RB2 test so as to get the fuel ejection, thereby obtaining information on the fuel ejection behavior....

  2. Funcionalidad de juegos de estrategia virtuales y del software Cabri-géomètre II en el aprendizaje de la simetría en secundaria

    OpenAIRE

    Guadalupe Rodríguez; Verónica Hoyos

    2010-01-01

    Presentamos resultados de un estudio exploratorio cuyo propósito es indagar sobre el uso del Cabri-Géomètre II y de un juego matemático virtual de estrategia cuando se incluyen en clases ordinarias de matemáticas de estudiantes de 12-13 años en una escuela secundaria pública de México. Para este trabajo, el tema abordado fue la simetría. El análisis de datos permitió ver la funcionalidad de estos ambientes para transformar las nociones del alumno sobre el tema y para reconocer que se puede av...

  3. Introduction d'une vue textuelle synchronisée avec la vue géométrique primaire dans Cabri-II

    OpenAIRE

    Bellynck, Valérie

    1999-01-01

    Cabri-géomètre est un logiciel qui permet l'exploration de figures géométriques par manipulation directe des objets géométriques qui les constituent. Ce logiciel plonge l'utilisateur dans un micromonde intelligent et constitue ainsi un environnement d'apprentissage pour la géométrie. Les utilisateurs peuvent construire des figures géométriques, explorer le champ des animations et déformations de la construction, élaborer de nouveaux outils avec des macro-constructions, et spécialiser leur env...

  4. Un modelo propuesto a partir de la inteligencia artificial y la didáctica. El ejemplo de Cabri-Euclide

    OpenAIRE

    Luengo, Vanda

    2005-01-01

    En el presente artículo expondremos un modelo inspirado en la teoría Semi-empírica y en la teoría de situaciones. La teoría semi-empírica ha sido propuesta dentro del área de la inteligencia artificial y la teoría de situaciones pertenece al área de la didáctica. En la primera parte presentaremos estas dos teorías, para luego introducir nuestro modelo. Finalmente, ilustraremos este modelo con la presentación de un programa de aprendizaje de la prueba en geometría: Cabri- Euclide.

  5. RELAP5-3D Restart and Backup Verification Testing

    Energy Technology Data Exchange (ETDEWEB)

    Dr. George L Mesina

    2013-09-01

    Existing testing methodology for RELAP5-3D employs a set of test cases collected over two decades to test a variety of code features and run on a Linux or Windows platform. However, this set has numerous deficiencies in terms of code coverage, detail of comparison, running time, and testing fidelity of RELAP5-3D restart and backup capabilities. The test suite covers less than three quarters of the lines of code in the relap directory and just over half those in the environmental library. Even in terms of code features, many are not covered. Moreover, the test set runs many problems long past the point necessary to test the relevant features. It requires standard problems to run to completion. This is unnecessary for features can be tested in a short-running problem. For example, many trips and controls can be tested in the first few time steps, as can a number of fluid flow options. The testing system is also inaccurate. For the past decade, the diffem script has been the primary tool for checking that printouts from two different RELAP5-3D executables agree. This tool compares two output files to verify that all characters are the same except for those relating to date, time and a few other excluded items. The variable values printed on the output file are accurate to no more than eight decimal places. Therefore, calculations with errors in decimal places beyond those printed remain undetected. Finally, fidelity of restart is not tested except in the PVM sub-suite and backup is not specifically tested at all. When a restart is made from any midway point of the base-case transient, the restart must produce the same values. When a backup condition occurs, the code repeats advancements with the same time step. A perfect backup can be tested by forcing RELAP5 to perform a backup by falsely setting a backup condition flag at a user-specified-time. Comparison of the calculations of that run and those produced by the same input w/o the spurious condition should be

  6. RELAP5-3D Resolution of Known Restart/Backup Issues

    Energy Technology Data Exchange (ETDEWEB)

    Mesina, George L.; Anderson, Nolan A.

    2014-12-01

    The state-of-the-art nuclear reactor system safety analysis computer program developed at the Idaho National Laboratory (INL), RELAP5-3D, continues to adapt to changes in computer hardware and software and to develop to meet the ever-expanding needs of the nuclear industry. To continue at the forefront, code testing must evolve with both code and industry developments, and it must work correctly. To best ensure this, the processes of Software Verification and Validation (V&V) are applied. Verification compares coding against its documented algorithms and equations and compares its calculations against analytical solutions and the method of manufactured solutions. A form of this, sequential verification, checks code specifications against coding only when originally written then applies regression testing which compares code calculations between consecutive updates or versions on a set of test cases to check that the performance does not change. A sequential verification testing system was specially constructed for RELAP5-3D to both detect errors with extreme accuracy and cover all nuclear-plant-relevant code features. Detection is provided through a “verification file” that records double precision sums of key variables. Coverage is provided by a test suite of input decks that exercise code features and capabilities necessary to model a nuclear power plant. A matrix of test features and short-running cases that exercise them is presented. This testing system is used to test base cases (called null testing) as well as restart and backup cases. It can test RELAP5-3D performance in both standalone and coupled (through PVM to other codes) runs. Application of verification testing revealed numerous restart and backup issues in both standalone and couple modes. This document reports the resolution of these issues.

  7. Production of radionuclides in nuclear reactor

    International Nuclear Information System (INIS)

    Given is a short review on the production of radionuclides which was performed in the Vinca Institute of Nuclear Sciences by using the nuclear reactor RA. Regarding the considerations of the possible re-starting of this reactor its use for the production of medical radionuclides should be taken into account. Listed are some of the important medical radionuclides routinely produced in nuclear reactors in the world and discussed the conditions for their obtaining in the reactor RA. (author)

  8. Funcionalidad de Juegos de Estrategia Virtuales y del Software Cabri-géomètre II en el Aprendizaje de la Simetría en Secundaria

    Directory of Open Access Journals (Sweden)

    Guadalupe Rodríguez

    2010-06-01

    Full Text Available Presentamos resultados de un estudio exploratorio cuyo propósito es indagar sobre el uso del Cabri-Géomètre II y de un juego matemático virtual de estrategia cuando se incluyen en clases ordinarias de matemáticas de estudiantes de 12-13 años en una escuela secundaria pública de México. Para este trabajo, el tema abordado fue la simetría. El análisis de datos permitió ver la funcionalidad de estos ambientes para transformar las nociones del alumno sobre el tema y para reconocer que se puede avanzar en el desarrollo de un pensamiento matemático distinto a través de ambientes de aprendizaje computacionales. We present the results of an exploratory study whose purpose was to explore the use by 12-13 years old students of Cabri-Géomètre II and a mathematical strategic virtual game in a Mexican state school. For this paper, the mathematical focus was on symmetry. The data analysis allowed us to observe the functionality of this learning environment to transform the students’ conceptions and to recognize that a different type of mathematical thinking can be developed with these computational learning environments.

  9. Why Electricite de France wanted to re-start Creys-Malville?

    International Nuclear Information System (INIS)

    This paper discusses the reasons in favour of a re-start up of Superphenix reactor. Two main reasons are developed: the first is that nuclear industry must prepare its future and invest in its future tools, and the second is that in the complex cycle of nuclear power each way must be explored to maintain its soundness. The paper recalls the successive historical steps which have led to the shutdown of the reactor, but also to the successive improvements developed to increase its safety and performances. This testing phase, which must be considered as an absolute necessity before any industrial development, has led to several cost overruns which have been the target to public criticisms. The paper focusses also on the importance of Superphenix technology in plutonium recycle and actinides burning for the reduction of plutonium stocks and the elimination of long-life nuclear wastes. In conclusion, Superphenix should not be considered as a final goal but as an intermediate step and a useful experience for the development of future FBR technology. (J.S.)

  10. Berkeley lab checkpoint/restart (BLCR) for Linux clusters

    Science.gov (United States)

    Hargrove, Paul H.; Duell, Jason C.

    2006-09-01

    This article describes the motivation, design and implementation of Berkeley Lab Checkpoint/Restart (BLCR), a system-level checkpoint/restart implementation for Linux clusters that targets the space of typical High Performance Computing applications, including MPI. Application-level solutions, including both checkpointing and fault-tolerant algorithms, are recognized as more time and space efficient than system-level checkpoints, which cannot make use of any application-specific knowledge. However, system-level checkpointing allows for preemption, making it suitable for responding to ''fault precursors'' (for instance, elevated error rates from ECC memory or network CRCs, or elevated temperature from sensors). Preemption can also increase the efficiency of batch scheduling; for instance reducing idle cycles (by allowing for shutdown without any queue draining period or reallocation of resources to eliminate idle nodes when better fitting jobs are queued), and reducing the average queued time (by limiting large jobs to running during off-peak hours, without the need to limit the length of such jobs). Each of these potential uses makes BLCR a valuable tool for efficient resource management in Linux clusters.

  11. Berkeley lab checkpoint/restart (BLCR) for Linux clusters

    International Nuclear Information System (INIS)

    This article describes the motivation, design and implementation of Berkeley Lab Checkpoint/Restart (BLCR), a system-level checkpoint/restart implementation for Linux clusters that targets the space of typical High Performance Computing applications, including MPI. Application-level solutions, including both checkpointing and fault-tolerant algorithms, are recognized as more time and space efficient than system-level checkpoints, which cannot make use of any application-specific knowledge. However, system-level checkpointing allows for preemption, making it suitable for responding to ''fault precursors'' (for instance, elevated error rates from ECC memory or network CRCs, or elevated temperature from sensors). Preemption can also increase the efficiency of batch scheduling; for instance reducing idle cycles (by allowing for shutdown without any queue draining period or reallocation of resources to eliminate idle nodes when better fitting jobs are queued), and reducing the average queued time (by limiting large jobs to running during off-peak hours, without the need to limit the length of such jobs). Each of these potential uses makes BLCR a valuable tool for efficient resource management in Linux clusters

  12. DMTCP: bringing interactive checkpoint-restart to Python

    Science.gov (United States)

    Arya, Kapil; Cooperman, Gene

    2015-01-01

    DMTCP (Distributed MultiThreaded CheckPointing) is a mature checkpoint-restart package. It operates in user space without kernel privilege, and adapts to application-specific requirements through plugins. While DMTCP has been able to checkpoint Python and IPython ‘from the outside’ for many years, a Python module has recently been created to support DMTCP. IPython support is included through a new DMTCP plugin. A checkpoint can be requested interactively within a Python session or under the control of a specific Python program. Further, the Python program can execute specific Python code prior to checkpoint, upon resuming (within the original process) and upon restarting (from a checkpoint image). Applications of DMTCP are demonstrated for: (i) Python-based graphics using virtual network client, (ii) a fast/slow technique to use multiple hosts or cores to check one (Cython Behnel S et al 2011 Comput. Sci. Eng. 13 31-39) computation in parallel, and (iii) a reversible debugger, FReD, with a novel reverse-expression watchpoint feature for locating the cause of a bug.

  13. Friction Stir Weld Restart+Reweld Repair Allowables

    Science.gov (United States)

    Clifton, Andrew

    2008-01-01

    A friction stir weld (FSW) repair method has been developed and successfully implemented on Al 2195 plate material for the Space Shuttle External Fuel Tank (ET). The method includes restarting the friction stir weld in the termination hole of the original weld followed by two reweld passes. Room temperature and cryogenic temperature mechanical properties exceeded minimum FSW design strength and compared well with the development data. Simulated service test results also compared closely to historical data for initial FSW, confirming no change to the critical flaw size or inspection requirements for the repaired weld. Testing of VPPA fusion/FSW intersection weld specimens exhibited acceptable strength and exceeded the minimum design value. Porosity, when present at the intersection was on the root side toe of the fusion weld, the "worst case" being 0.7 inch long. While such porosity may be removed by sanding, this "worst case" porosity condition was tested "as is" and demonstrated that porosity did not negatively affect the strength of the intersection weld. Large, 15-inch "wide panels" FSW repair welds were tested to demonstrate strength and evaluate residual stresses using photo stress analysis. All results exceeded design minimums, and photo stress analysis showed no significant stress gradients due to the presence of the restart and multi-pass FSW repair weld.

  14. Asymptotic optimality of RESTART estimators in highly dependable systems

    International Nuclear Information System (INIS)

    We consider a wide class of models that includes the highly reliable Markovian systems (HRMS) often used to represent the evolution of multi-component systems in reliability settings. Repair times and component lifetimes are random variables that follow a general distribution, and the repair service adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly-dependable systems, the RESTART method is used for the estimation of steady-state unavailability and other reliability measures. In this method, a number of simulation retrials are performed when the process enters regions of the state space where the chance of occurrence of a rare event (e.g., a system failure) is higher. The main difficulty involved in applying this method is finding a suitable function, called the importance function, to define the regions. In this paper we introduce an importance function which, for unbalanced systems, represents a great improvement over the importance function used in previous papers. We also demonstrate the asymptotic optimality of RESTART estimators in these models. Several examples are presented to show the effectiveness of the new approach, and probabilities up to the order of 10−42 are accurately estimated with little computational effort. - Highlights: • Rare event probabilities of highly reliable systems are estimated by simulation. • The asymptotic optimality of the application is proved. • A better importance function for highly reliable systems is provided in the paper

  15. Keeping checkpoint/restart viable for exascale systems.

    Energy Technology Data Exchange (ETDEWEB)

    Riesen, Rolf E.; Bridges, Patrick G. (IBM Research, Ireland, Mulhuddart, Dublin); Stearley, Jon R.; Laros, James H., III; Oldfield, Ron A.; Arnold, Dorian (University of New Mexico, Albuquerque, NM); Pedretti, Kevin Thomas Tauke; Ferreira, Kurt Brian; Brightwell, Ronald Brian

    2011-09-01

    Next-generation exascale systems, those capable of performing a quintillion (10{sup 18}) operations per second, are expected to be delivered in the next 8-10 years. These systems, which will be 1,000 times faster than current systems, will be of unprecedented scale. As these systems continue to grow in size, faults will become increasingly common, even over the course of small calculations. Therefore, issues such as fault tolerance and reliability will limit application scalability. Current techniques to ensure progress across faults like checkpoint/restart, the dominant fault tolerance mechanism for the last 25 years, are increasingly problematic at the scales of future systems due to their excessive overheads. In this work, we evaluate a number of techniques to decrease the overhead of checkpoint/restart and keep this method viable for future exascale systems. More specifically, this work evaluates state-machine replication to dramatically increase the checkpoint interval (the time between successive checkpoint) and hash-based, probabilistic incremental checkpointing using graphics processing units to decrease the checkpoint commit time (the time to save one checkpoint). Using a combination of empirical analysis, modeling, and simulation, we study the costs and benefits of these approaches on a wide range of parameters. These results, which cover of number of high-performance computing capability workloads, different failure distributions, hardware mean time to failures, and I/O bandwidths, show the potential benefits of these techniques for meeting the reliability demands of future exascale platforms.

  16. Reactors

    International Nuclear Information System (INIS)

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

  17. Stop and Restart Effects on Modern Vehicle Starting System Components

    Energy Technology Data Exchange (ETDEWEB)

    Windover, Paul R. [Argonne National Lab. (ANL), Argonne, IL (United States); Owens, Russell J. [Argonne National Lab. (ANL), Argonne, IL (United States); Levinson, Terry M. [Argonne National Lab. (ANL), Argonne, IL (United States); Laughlin, Michael [Argonne National Lab. (ANL), Argonne, IL (United States); Gaines, Linda [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-01

    Many drivers of personal and commercial vehicles believe that turning the vehicle off and on frequently instead of idling will cause premature wear of the starter system (starter motor and starter battery). As a result, they are concerned that the replacement cost of the starter motor and/or battery due to increased manual engine cycling would be more than the cumulative cost of the fuel saved by not idling unnecessarily. A number of variables play a role in addressing this complex concern, including the number of starting cycles per day, the time between starting cycles, the intended design life of the starting system, the amount of fuel used to restart an engine, and the cumulative cost of the saved fuel. Qualitative and quantitative information from a variety of sources was used to develop a life-cycle economic model to evaluate the cost and quantify the realistic factors that are related to the permissible frequency of starter motor cycles for the average vehicle to economically minimize engine idle time. Annual cost savings can be calculated depending on shutdown duration and the number of shutdown cycles per day. Analysis shows that cost savings are realized by eliminating idling exceeding one minute by shutting down the engine and restarting it. For a typical motorist, the damage to starting system components resulting from additional daily start cycles will be negligible. Overall, it was found that starter life is mostly dependent on the total number of start cycles, while battery life is more dependent on ensuring a full charge between start events.

  18. The Cabri installation - INB 24. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    After a presentation of some characteristics of the Cabri base nuclear installation located in Cadarache (brief description, buildings, operation), this document reports the identification of cliff-edge effect risks and of critical structures and equipment. Then, it addresses the different risks: earthquake (installation sizing and compliance, margin assessment for the different structures and equipment), external flooding (installation sizing and compliance, margin assessment in relationship with the different flooding origins), other extreme natural events (extreme meteorological conditions related to flooding, earthquake exceeding the design level), and loss of external or internal electric supplies and of cooling systems. The next parts address severe accident management (means and organization for crisis management, robustness of available means), and subcontracting conditions and practices

  19. Proper disposal of irradiated fuel as the first step towards deciding the RA research reactor future

    International Nuclear Information System (INIS)

    The first step towards restarting, or eventual decommissioning, of the RA research reactor must be safe and reliable disposal of spent fuel. Adequate storage of fuel irradiated so far is to be provided, as well as disposal of new irradiated fuel if or when the reactor is restarted. Activities presented here are related to identification and minimization of corrosion processes and further degradation in the existing storage pool. In the second phase, a new, presumably dry storage facility should be built. (author)

  20. FBR type reactor

    International Nuclear Information System (INIS)

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

  1. Modelling Validation of Transients and Initial Phase of Accident Scenarios for Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Physical phenomena are presented being of importance in case of transients and / or initial phases of severe accidents in Sodium-cooled Fast Reactors. The CABRI-programmes provided experimental data being characteristic for the physical phenomena and providing information to validate models and parameters used in theoretical simulations. Results of post irradiation examination (PIE), post test examination (PTE) and measurements performed during the experimental tests are presented for transient overpower (TOP), transient undercooling overpower (TUCOP), loss-of-flow tests (LOF) and slow power ramps. (author)

  2. Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System n removed by the passive Reactor (SCS) are assumed to have failed. Decay heat is the Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

  3. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.

  4. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-09-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.

  5. Clean energy for a new generation. Steam generator life cycle management and Bruce restart

    International Nuclear Information System (INIS)

    In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre

  6. Latest news from the YETS: all restarting except the LHC

    CERN Multimedia

    Antonella Del Rosso

    2016-01-01

    With the closure of the SPS at 3 p.m. on Friday, 19 February, maintenance work is now ongoing only at the LHC. All activities are on track for a smooth restart in a few weeks’ time.   At the LHC, all general maintenance activities are proceeding well and according to schedule. In particular, the electrical tests on the general emergency stops have been completed, while the cooling and ventilation maintenance, including leak repairs at various points, will be completed by the end of this week for the whole machine. By the end of next week, the teams will also have completed the bakeout and commissioning of all the collimators, while the installation of coaxial cable for clock distribution for CMS TOTEM has been postponed to the Extended Year-End Technical Stop (EYETS), scheduled to start in December. Following a recent decision, additional electrical tests of the circuits (ELQA and energy extraction insulation tests) for the whole machine have been added to the schedule. The SPS is currentl...

  7. Markov Renewal Methods in Restart Problems in Complex Systems

    DEFF Research Database (Denmark)

    Asmussen, Søren; Lipsky, Lester; Thompson, Stephen

    general alternating Markov renewal model is proposed and an asymptotic exponential form P(X > x) ∼ Ce−γx identified for the case of a deterministic task time L ≡ `. The rate γ is given by equating the spectral radius of a certain matrix to 1, and the asymptotic form of γ = γ(`) as ` → ∞ is derived......A task with ideal execution time L such as the execution of a computer program or the transmission of a file on a data link may fail, and the task then needs to be restarted. The task is handled by a complex system with features similar to the ones in classical reliability: failures may be......, leading to the asymptotics of P(X > x) for random task times L. A main finding is that X is always heavy-tailed if L has unbounded support. The case where the Markov renewal model is derived by lumping in a continuous-time finite Markov process with exponential holding times is given special attention...

  8. Optimal Rapid Restart of Heuristic Methods of NP Hard Problems

    Institute of Scientific and Technical Information of China (English)

    侯越先; 王芳

    2004-01-01

    Many heuristic search methods exhibit a remarkable variability in the time required to solve some particular problem instances. Their cost distributions are often heavy-tailed. It has been demonstrated that, in most cases, rapid restart (RR) method can prominently suppress the heavy-tailed nature of the instances and improve computation efficiency. However, it is usually time-consuming to check whether an algorithm on a specific instance is heavy-tailed or not. Moreover, if the heavy-tailed distribution is confirmed and the RR method is relevant, an optimal RR threshold should be chosen to facilitate the RR mechanism. In this paper, an approximate approach is proposed to quickly check whether an algorithm on a specific instance is heavy-tailed or not.The method is realized by means of calculating the maximal Lyapunov exponent of its generic running trace.Then a statistical formula to estimate the optimal RR threshold is educed. The method is based on common nonparametric estimation, e. g. , Kernel estimation. Two heuristic methods are selected to verify our method. The experimental results are consistent with the theoretical consideration perfectly.

  9. Restart oversight assessment of Hanford 242-A evaporator: Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Lagdon, R.; Lasky, R.

    1994-08-01

    An assessment team from the Office of Environment, Safety and Health (EH), US Department of Energy (DOE), conducted an independent assessment of the 242-A Evaporator at the Hanford Site during January 17--28, 1994. An EH team member remained on-site following the assessment to track corrective actions and resolve prestart findings. The primary objective of this assessment was independent assurance that the DOE Office of Environmental Management (EM), the DOE Richland Operations Office (DOE-RL), and Westinghouse Hanford Company (WHC) can safely restart the evaporator. Another objective of the EH team was to assess EM`s Operational Readiness Evaluation (ORE) to determine if the programs, procedures, and management systems implemented for operation of the 241-A Evaporator ensure the protection of worker safety and health. The following section of this report provides background information on the 242-A Evaporator and Operational Readiness Review (ORR) activities conducted to date. The next chapter is divided into sections that address the results of discrete assessment activities. Each section includes a brief statement of conclusions for the functional area in question, descriptions of the review bases and methods, and a detailed discussion of the results. Concerns identified during the assessment are listed for the section to which they apply, and the specific findings upon which the concern is based can be found immediately thereafter.

  10. Restart oversight assessment of Hanford 242-A evaporator: Technical report

    International Nuclear Information System (INIS)

    An assessment team from the Office of Environment, Safety and Health (EH), US Department of Energy (DOE), conducted an independent assessment of the 242-A Evaporator at the Hanford Site during January 17--28, 1994. An EH team member remained on-site following the assessment to track corrective actions and resolve prestart findings. The primary objective of this assessment was independent assurance that the DOE Office of Environmental Management (EM), the DOE Richland Operations Office (DOE-RL), and Westinghouse Hanford Company (WHC) can safely restart the evaporator. Another objective of the EH team was to assess EM's Operational Readiness Evaluation (ORE) to determine if the programs, procedures, and management systems implemented for operation of the 241-A Evaporator ensure the protection of worker safety and health. The following section of this report provides background information on the 242-A Evaporator and Operational Readiness Review (ORR) activities conducted to date. The next chapter is divided into sections that address the results of discrete assessment activities. Each section includes a brief statement of conclusions for the functional area in question, descriptions of the review bases and methods, and a detailed discussion of the results. Concerns identified during the assessment are listed for the section to which they apply, and the specific findings upon which the concern is based can be found immediately thereafter

  11. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  12. ESP – Data from Restarted Life Tests of Various Silicon Materials

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Jim

    2010-10-06

    Current funding has allowed the restart of testing of various silicone materials placed in Life Tests or Aging Studies from past efforts. Some of these materials have been in test since 1982, with no testing for approximately 10 years, until funding allowed the restart in FY97. Charts for the various materials at different thickness, compression, and temperature combinations illustrate trends for the load-bearing properties of the materials.

  13. Some comparison of restarted GMRES and QMR for linear and nonlinear problems

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, R. [Baylor Univ., Waco, TX (United States); Joubert, W. [Los Alamos National Lab., NM (United States)

    1994-12-31

    Comparisons are made between the following methods: QMR including its transpose-free version, restarted GMRES, and a modified restarted GMRES that uses approximate eigenvectors to improve convergence, For some problems, the modified GMRES is competitive with or better than QMR in terms of the number of matrix-vector products. Also, the GMRES methods can be much better when several similar systems of linear equations must be solved, as in the case of nonlinear problems and ODE problems.

  14. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

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

  15. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

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

  16. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

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

  17. Environmental Information Document: L-reactor reactivation

    International Nuclear Information System (INIS)

    Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program

  18. Environmental Information Document: L-reactor reactivation

    Energy Technology Data Exchange (ETDEWEB)

    Mackey, H.E. Jr. (comp.)

    1982-04-01

    Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

  19. RA reactor operation, Annex A

    International Nuclear Information System (INIS)

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

  20. Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, A.F.; Spangler, J.B.

    1995-04-05

    An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH& QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment.

  1. Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

    International Nuclear Information System (INIS)

    An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH ampersand QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment

  2. Benchmarking the Nelder-Mead Downhill Simplex Algorithm With Many Local Restarts

    OpenAIRE

    Hansen, Nikolaus

    2009-01-01

    International audience We benchmark the Nelder-Mead downhill simplex method on the noisefree BBOB-2009 testbed. A multistart strategy is applied on two levels. On a local level, at least ten restarts are conducted with a small number of iterations and reshaped simplex. On the global level independent restarts are launched until $10^5 D$ function evaluations are exceeded, for dimension $D\\ge20$ ten times less. For low search space dimensions the algorithm shows very good results on many fun...

  3. Increased capability gas generator for Space Shuttle APU. Development/hot restart test report

    Science.gov (United States)

    1980-01-01

    The design, fabrication, and testing of an increased capability gas generator for use in space shuttles are described. Results show an unlimited hot restart capability in the range of feed pressures from 400 psi to 80 psi. Effects of vacuum on hot restart were not addressed, and only beginning-of-life bed conditions were tested. No starts with bubbles were performed. A minimum expected life of 35 hours or more is projected, and the design will maintain a surface temperature of 350 F or more.

  4. Development and Analysis of the FlowManager? Cooldown and Restart Simulator

    OpenAIRE

    Solvang, Bård

    2008-01-01

    During the work presented in this thesis, the Shut-down flow model of the FlowManager? Cooldown and Restart Simulator has been developed into a transient multiphase flow model that is able to handle shut-down flow and restart flow for horizontal pipelines with uniform discretization.It has been shown that the momentum equation solver originally used in the Shut-down flow model fails to transfer momentum across velocity sign changes. As a result of this, the original momentum equation solver m...

  5. An Investigation of Restarted GMRES Method by Using Flexible Starting Vectors

    Institute of Scientific and Technical Information of China (English)

    Qiang Niu; Lin-Zhang Lu

    2010-01-01

    We discuss a variant of restarted GMRES method that allows changes of the restarting vector at each cycle of iterations. The merit of the variant is that previously generated information can be utilized to select a new starting vector, such that the occurrence of stagnation be mitigated or the convergence be accelerated. The more appealing utilization of the new method is in conjunction with a harmonic Ritz vector as the starting vector, which is discussed in detail. Numerical experiments are carried out to demonstrate that the proposed procedure can effectively mitigate the occurrence of stagnation due to the presence of small eigenvalues in modulus.

  6. Aplicación de la metodología interactiva del dibujo técnico en la enseñanza secundaria con el programa CABRI 2D-3D

    OpenAIRE

    TORRES BUITRAGO, RAFAEL JESÚS

    2010-01-01

    La investigación parte de las dificultades y problemas detectados en el aprendizaje del Dibujo Técnico en los alumnos de bachillerato dentro del aula. Desde este punto de partida nos planteamos, intentar mejorar la calidad de la enseñanza, ayudando a los alumnos a desarrollar la capacidad espacial y el razonamiento abstracto con un aprendizaje no memorístico aportando una nueva metodología educativa, basada en el empleo de programas informáticos (procesadores matemáticos) como Cabri 2D-3D, ...

  7. La transformación de rotación en el espacio: diseño curricular e integración en el aula del ambiente de geometría dinámica Cabri 3D

    OpenAIRE

    Alvarez, Zaira; Fernández, Daniel

    2008-01-01

    Se busca generar una discusión sobre el proceso de diseño y sistematización de una experiencia de aula en la cual se integra el Ambiente de Geometría Dinámica (AGD) Cabri 3D en el aprendizaje de la transformación de rotación en el espacio. En nuestra propuesta, encontramos investigaciones importantes en didáctica de las matemáticas que han puesto en evidencia las dificultades que los estudiantes presentan comúnmente en la exploración de propiedades de los objetos geométricos en el espacio...

  8. Utilización del programa Cabri 3D como herramienta didáctica para la enseñanza de Geometría en 2º de ESO

    OpenAIRE

    Martín-García, Noelia

    2013-01-01

    Este trabajo tiene como objetivo principal el planteamiento de una propuesta didáctica para la utilización del software Cabri 3D como recurso didáctico para la enseñanza de Geometría en la asignatura de Matemáticas, más concretamente en 2º de la ESO, para fomentar el aprendizaje significativo del alumnado. Para ello se ha realizado tanto una investigación bibliográfica como un estudio de campo. Durante la investigación bibliográfica se han consultado diferentes informes, artícu...

  9. Control Rod Worth Measurement in Monju Restart Core

    International Nuclear Information System (INIS)

    The Japanese prototype fast breeder reactor (FBR) Monju resumed the system startup test (SST) in May 2010 after fourteen year and five month suspension since the sodium leakage of the secondary heat transport system in December 1995. Core confirmation test (CCT) is the first step of SST which consists of three steps, and finished in July 2010. Valuable basic data for FBR development was obtained in CCT, such as reactor physics data of the core which contained 1.5wt%/HM in average of Am-241 accumulated due to the Pu-241 decay during the long-term suspension. Control rod reactivity worth measurement was carried out to calibrate the reactivity worth of control rods and to confirm the core characteristics such as excess reactivity and reactivity shutdown margin to be satisfied with safety criteria. The high prediction accuracy of the core management code system was demonstrated based on the measured data. Furthermore, the examination was conducted to shorten the measurement period. (author)

  10. FBR type reactor

    International Nuclear Information System (INIS)

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

  11. Initialization and Restart in Stochastic Local Search: Computing a Most Probable Explanation in Bayesian Networks

    Science.gov (United States)

    Mengshoel, Ole J.; Wilkins, David C.; Roth, Dan

    2010-01-01

    For hard computational problems, stochastic local search has proven to be a competitive approach to finding optimal or approximately optimal problem solutions. Two key research questions for stochastic local search algorithms are: Which algorithms are effective for initialization? When should the search process be restarted? In the present work we investigate these research questions in the context of approximate computation of most probable explanations (MPEs) in Bayesian networks (BNs). We introduce a novel approach, based on the Viterbi algorithm, to explanation initialization in BNs. While the Viterbi algorithm works on sequences and trees, our approach works on BNs with arbitrary topologies. We also give a novel formalization of stochastic local search, with focus on initialization and restart, using probability theory and mixture models. Experimentally, we apply our methods to the problem of MPE computation, using a stochastic local search algorithm known as Stochastic Greedy Search. By carefully optimizing both initialization and restart, we reduce the MPE search time for application BNs by several orders of magnitude compared to using uniform at random initialization without restart. On several BNs from applications, the performance of Stochastic Greedy Search is competitive with clique tree clustering, a state-of-the-art exact algorithm used for MPE computation in BNs.

  12. Restart Plan for the Prototype Vertical Denitration Calciner [SD Coversheet has Incorrect Document Number

    Energy Technology Data Exchange (ETDEWEB)

    SUTTER, C.S.

    1999-07-26

    Testing activities on the Prototype Vertical Denitration Calciner at PFP were suspended in January 1997 due to the hold on fissile material handling in the facility. The Restart Plan will govern the transition of the test program from the completion of the activity based startup review; through equipment checkout and surrogate material runs; to resumption of the testing program and transition to unrestricted testing.

  13. Removal of floating organic in Hanford Waste Tank 241-C-103 restart plan

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, T.R.; Hanson, C.

    1994-10-03

    The decision whether or not to remove the organic layer from Waste Tank 241-C-103 was deferred until May, 1995. The following restart plan was prepared for removal of the organic if the decision is to remove the organic from the waste tank 241-C-103.

  14. Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

    International Nuclear Information System (INIS)

    In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements

  15. Removal of floating organic in Hanford Waste Tank 241-C-103 restart plan

    International Nuclear Information System (INIS)

    The decision whether or not to remove the organic layer from Waste Tank 241-C-103 was deferred until May, 1995. The following restart plan was prepared for removal of the organic if the decision is to remove the organic from the waste tank 241-C-103

  16. RA reactor operation and maintenance in 1999, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1999 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  17. RA reactor operation and maintenance in 1998, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1998 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  18. Validation of DRAGON side-step method for Bruce-A restart Phase-B physics tests

    International Nuclear Information System (INIS)

    The DRAGON side-step method, developed at AECL, has a number of advantages over the all-DRAGON method that was used before. It is now the qualified method for reactivity-device calculations. Although the side-step-method-generated incremental cross sections have been validated against those previously calculated with the all-DRAGON method, it is highly desirable to validate the side-step method against device-worth measurements in power reactors directly. In this paper, the DRAGON side-step method was validated by comparison with the device-calibration measurements made in Bruce-A NGS Unit 4 restart Phase-B commissioning in 2003. The validation exercise showed excellent results, with the DRAGON code overestimating the measured ZCR worth by ∼5%. A sensitivity study was also performed in this paper to assess the effect of various DRAGON modelling techniques on the incremental cross sections. The assessment shows that the refinement of meshes in 3-D and the use of the side-step method are two major reasons contributing to the improved agreement between the calculated ZCR worths and the measurements. Use of different DRAGON versions, DRAGON libraries, local-parameter core conditions, and weighting techniques for the homogenization of tube clusters inside the ZCR have a very small effect on the ZCR incremental thermal absorption cross section and ZCR reactivity worth. (author)

  19. 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

  20. Design study of eventual core conversion for the research reactor RA

    International Nuclear Information System (INIS)

    Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

  1. Deflated and restarted symmetric Lanczos methods for eigenvalues and linear equations with multiple right-hand sides

    CERN Document Server

    Abdel-Rehim, Abdou M; Nicely, Dywayne A; Wilcox, Walter

    2008-01-01

    A deflated restarted Lanczos algorithm is given for both solving symmetric linear equations and computing eigenvalues and eigenvectors. The restarting limits the storage so that finding eigenvectors is practical. Meanwhile, the deflating from the presence of the eigenvectors allows the linear equations to generally have good convergence in spite of the restarting. Some reorthogonalization is necessary to control roundoff error, and several approaches are discussed. The eigenvectors generated while solving the linear equations can be used to help solve systems with multiple right-hand sides. Experiments are given with large matrices from quantum chromodynamics that have many right-hand sides.

  2. Naturalistic field study of the restart break in US commercial motor vehicle drivers: Truck driving, sleep, and fatigue.

    Science.gov (United States)

    Sparrow, Amy R; Mollicone, Daniel J; Kan, Kevin; Bartels, Rachel; Satterfield, Brieann C; Riedy, Samantha M; Unice, Aaron; Van Dongen, Hans P A

    2016-08-01

    Commercial motor vehicle (CMV) drivers in the US may start a new duty cycle after taking a 34-h restart break. A restart break provides an opportunity for sleep recuperation to help prevent the build-up of fatigue across duty cycles. However, the effectiveness of a restart break may depend on its timing, and on how many nighttime opportunities for sleep it contains. For daytime drivers, a 34-h restart break automatically includes two nighttime periods. For nighttime drivers, who are arguably at increased risk of fatigue, a 34-h restart break contains only one nighttime period. To what extent this is relevant for fatigue depends in part on whether nighttime drivers revert back to a nighttime-oriented sleep schedule during the restart break. We conducted a naturalistic field study with 106 CMV drivers working their normal schedules and performing their normal duties. These drivers were studied during two duty cycles and during the intervening restart break. They provided a total of 1260days of data and drove a total of 414,937 miles during the study. Their duty logs were used to identify the periods when they were on duty and when they were driving and to determine their duty cycles and restart breaks. Sleep/wake patterns were measured continuously by means of wrist actigraphy. Fatigue was assessed three times per day by means of a brief psychomotor vigilance test (PVT-B) and a subjective sleepiness scale. Data from a truck-based lane tracking and data acquisition system were used to compute lane deviation (variability in lateral lane position). Statistical analyses focused on 24-h patterns of duty, driving, sleep, PVT-B performance, subjective sleepiness, and lane deviation. Duty cycles preceded by a restart break containing only one nighttime period (defined as 01:00-05:00) were compared with duty cycles preceded by a restart break containing more than one nighttime period. During duty cycles preceded by a restart break with only one nighttime period, drivers

  3. TMI-1 restart: an evaluation of the licensee's management integrity as it affects restart of Three Mile Island Nuclear Station (Unit 1 Docket 50-289). Supplement 5

    International Nuclear Information System (INIS)

    Supplement 5 to the Safety Evaluation Report (SER) on TMI-1 Restart documents the review by the Nuclear Regulatory Commission (NRC) staff of nine investigations conducted by the NRC Office of Investigations into matters identified as relevant and material to an evaluation of the licensee's management integrity. The staff has included, as part of its evaluation, materials from its review of the GPU v. B and W lawsuit record (NUREG-1020LD, GPU, v. B and W Lawsuit Review and Its Effect on TMI-1) as well as other relevant materials developed since the close of the record in the TMI-1 Restart proceeding. In developing its position on General Public Utilities Nuclear Corporation's character (i.e., management integrity), the staff evaluated matters that cast doubt on the licensee's character, individually and collectively; considered the remedial actions taken by the licensee; and balanced past improper conduct of the licensee against its subsequent record of remedial actions and performance and record of current senior management of the licensee. The staff concluded that, while the past improper conduct was grave, the remedial actions taken, the subsequent record of performance, and the record of current senior management support a finding that GPUN can and will operate TMI-1 without undue risk to the health and safety of the public

  4. Use of checkpoint-restart for complex HEP software on traditional architectures and Intel MIC

    International Nuclear Information System (INIS)

    Process checkpoint-restart is a technology with great potential for use in HEP workflows. Use cases include debugging, reducing the startup time of applications both in offline batch jobs and the High Level Trigger, permitting job preemption in environments where spare CPU cycles are being used opportunistically and efficient scheduling of a mix of multicore and single-threaded jobs. We report on tests of checkpoint-restart technology using CMS software, Geant4-MT (multi-threaded Geant4), and the DMTCP (Distributed Multithreaded Checkpointing) package. We analyze both single- and multi-threaded applications and test on both standard Intel x86 architectures and on Intel MIC. The tests with multi-threaded applications on Intel MIC are used to consider scalability and performance. These are considered an indicator of what the future may hold for many-core computing

  5. Refurbishment status on reactor facilities of JMTR

    International Nuclear Information System (INIS)

    The JMTR (Japan Materials Testing Reactor), a light-water-cooling tank-type reactor with a 50 MW thermal power, was shutdown in August 2006. The reactor facilities are to be refurbished during four years from the beginning of FY 2007, and the renewed JMTR will restart from FY 2011. In advance of the reactor refurbishment, equipments on reactor facilities to be renewed and to be continuously used were selected from a viewpoint of ensuring safety, improvement of operating availability, etc. The selected equipments to be renewed were the reactor instrument and control system, cooling system, radioactive waste facility, power supply system, boiler, etc. This report describes the basic idea on selection of the renewal facilities and schedule of refurbishment work. (author)

  6. Wind Turbine Shut Down and Restart Operations – A Test on 800kW WECS

    OpenAIRE

    Siddharth Sanjaykumar Joshi

    2011-01-01

    With the growing energy demand and need todecrease greenhouse gas emissions there has been a rise in thepopularity of renewable energy systems. One of the most popularrenewable energy systems over the past decade has been the windturbine. Technological advances in modeling can satisfy the needand interest of researcher. Present work give focus on windturbine shut downs and restart operations with its dominatingeffects on entire WECS starting from wind turbine to 33kV grid.Entire work has been...

  7. The design and implementation of Berkeley Lab's linuxcheckpoint/restart

    Energy Technology Data Exchange (ETDEWEB)

    Duell, Jason

    2005-04-30

    This paper describes Berkeley Linux Checkpoint/Restart (BLCR), a linux kernel module that allows system-level checkpoints on a variety of Linux systems. BLCR can be used either as a stand alone system for checkpointing applications on a single machine, or as a component by a scheduling system or parallel communication library for checkpointing and restoring parallel jobs running on multiple machines. Integration with Message Passing Interface (MPI) and other parallel systems is described.

  8. RESTART Simulation of Non-Markov Consecutive-K-Out-of-N: F Repairable Systems

    OpenAIRE

    Villén Altamirano, José

    2010-01-01

    The reliability of consecutive-k-out-of-n: F repairable systems and (k−1)-step Markov dependence is studied. The model analyzed in this paper is more general than those of previous studies given that repair time and component lifetimes are random variables that follow a general distribution. The system has one repair service which adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly dependable systems, the RESTART method wa...

  9. Data assimilation for unsaturated flow models with restart adaptive probabilistic collocation based Kalman filter

    Science.gov (United States)

    Man, Jun; Li, Weixuan; Zeng, Lingzao; Wu, Laosheng

    2016-06-01

    The ensemble Kalman filter (EnKF) has gained popularity in hydrological data assimilation problems. As a Monte Carlo based method, a sufficiently large ensemble size is usually required to guarantee the accuracy. As an alternative approach, the probabilistic collocation based Kalman filter (PCKF) employs the polynomial chaos expansion (PCE) to represent and propagate the uncertainties in parameters and states. However, PCKF suffers from the so-called "curse of dimensionality". Its computational cost increases drastically with the increasing number of parameters and system nonlinearity. Furthermore, PCKF may fail to provide accurate estimations due to the joint updating scheme for strongly nonlinear models. Motivated by recent developments in uncertainty quantification and EnKF, we propose a restart adaptive probabilistic collocation based Kalman filter (RAPCKF) for data assimilation in unsaturated flow problems. During the implementation of RAPCKF, the important parameters are identified and active PCE basis functions are adaptively selected at each assimilation step; the "restart" scheme is utilized to eliminate the inconsistency between updated model parameters and states variables. The performance of RAPCKF is systematically tested with numerical cases of unsaturated flow models. It is shown that the adaptive approach and restart scheme can significantly improve the performance of PCKF. Moreover, RAPCKF has been demonstrated to be more efficient than EnKF with the same computational cost.

  10. Michaelis-Menten reaction scheme as a unified approach towards the optimal restart problem

    Science.gov (United States)

    Rotbart, Tal; Reuveni, Shlomi; Urbakh, Michael

    2015-12-01

    We study the effect of restart, and retry, on the mean completion time of a generic process. The need to do so arises in various branches of the sciences and we show that it can naturally be addressed by taking advantage of the classical reaction scheme of Michaelis and Menten. Stopping a process in its midst—only to start it all over again—may prolong, leave unchanged, or even shorten the time taken for its completion. Here we are interested in the optimal restart problem, i.e., in finding a restart rate which brings the mean completion time of a process to a minimum. We derive the governing equation for this problem and show that it is exactly solvable in cases of particular interest. We then continue to discover regimes at which solutions to the problem take on universal, details independent forms which further give rise to optimal scaling laws. The formalism we develop, and the results obtained, can be utilized when optimizing stochastic search processes and randomized computer algorithms. An immediate connection with kinetic proofreading is also noted and discussed.

  11. An asynchronous writing method for restart files in the gysela code in prevision of exascale systems*

    Directory of Open Access Journals (Sweden)

    Thomine O.

    2013-12-01

    Full Text Available The present work deals with an optimization procedure developed in the full-f global GYrokinetic SEmi-LAgrangian code (GYSELA. Optimizing the writing of the restart files is necessary to reduce the computing impact of crashes. These files require a very large memory space, and particularly so for very large mesh sizes. The limited bandwidth of the data pipe between the comput- ing nodes and the storage system induces a non-scalable part in the GYSELA code, which increases with the mesh size. Indeed the transfer time of RAM to data depends linearly on the files size. The necessity of non synchronized writing-in-file procedure is therefore crucial. A new GYSELA module has been developed. This asynchronous procedure allows the frequent writ- ing of the restart files, whilst preventing a severe slowing down due to the limited writing bandwidth. This method has been improved to generate a checksum control of the restart files, and automatically rerun the code in case of a crash for any cause.

  12. Michaelis-Menten reaction scheme as a unified approach towards the optimal restart problem.

    Science.gov (United States)

    Rotbart, Tal; Reuveni, Shlomi; Urbakh, Michael

    2015-12-01

    We study the effect of restart, and retry, on the mean completion time of a generic process. The need to do so arises in various branches of the sciences and we show that it can naturally be addressed by taking advantage of the classical reaction scheme of Michaelis and Menten. Stopping a process in its midst-only to start it all over again-may prolong, leave unchanged, or even shorten the time taken for its completion. Here we are interested in the optimal restart problem, i.e., in finding a restart rate which brings the mean completion time of a process to a minimum. We derive the governing equation for this problem and show that it is exactly solvable in cases of particular interest. We then continue to discover regimes at which solutions to the problem take on universal, details independent forms which further give rise to optimal scaling laws. The formalism we develop, and the results obtained, can be utilized when optimizing stochastic search processes and randomized computer algorithms. An immediate connection with kinetic proofreading is also noted and discussed. PMID:26764608

  13. RA nuclear reactor - revitalisation, renewal and applications

    International Nuclear Information System (INIS)

    This book is meant to give professional support in solving the problem of RA reactor, its revitalisation and renewal, as a special help for decision makers. Facts in favor of restarting RA reactor are prevailing. This report is made of six parts. First part includes an overview of basic properties of research reactors in the world and a discussion concerning their future development. RA reactor parameters are analyzed both with low enriched and highly enriched fuel and it has been concluded that the aim of RA reactor renewal should be to obtaining as high as possible thermal neutron flux density. The second part deals with possible applications of RA reactor in fundamental and applied research programs, commercial applications and its role in education and training programs. The third part discusses application of RA reactor as a source of thermal neutrons for fundamental and applied sciences, especially in the condensed matter physics and development of new materials. The role of RA reactor in development of radiation protection systems is emphasised in part four. Some possible commercial applications of Ra reactor are described in part five: isotope production, and their different applications. Part six deals with education and training of staff, with special accent on scientific international cooperation. Basic conclusions of this material meant for decision makers are: restarting RA reactor is the most reasonable and activities related to its revitalisation and renewal should be continued; this program should include solving the problems of education and training of the staff for reactor operation, improvement and different applications; renewal program should include renewal of the experimental devices as a condition of reactor efficient application immediately after its startup

  14. Enfoque didáctico para la conceptualización de la parábola como lugar geométrico integrando Cabri Géomètre II Plus

    OpenAIRE

    Moncayo, Claudia Andrea; Pantoja, José Luis; Mosquera, Edinsson Fernández

    2012-01-01

    La enseñanza de la parábola suele restringirse al enfoque de la geometría analítica. El presente estudio consistió en una propuesta didáctica que acercó a los estudiantes a la comprensión del significado de parábola como lugar geométrico mediante el diseño e implementación de una estrategia didáctica basada en el uso e integración del ambiente de geometría dinámica Cabri Géomètre II Plus. En su diseño, puesta en práctica y sistematización, se consideraron las fases de una micro-ingeniería did...

  15. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  16. JMTR strategy of restart and dosimetry for standardization of irradiation technology

    International Nuclear Information System (INIS)

    Calculated neutron flux/fluence was verified against measurements of irradiated fluence monitors. With regard to gamma dose, calculated gamma heating rate were verified against measurements of the nuclear heating evaluation capsule which was developed in order to measure nuclear heating rate. It was confirmed that the calculated fast and thermal neutron flux/fluence agreed with measurements within ±10%, ±30%, respectively, and the calculated gamma dose agreed within -3 - ±21%. The attempt to improve accuracy of calculated irradiation parameter, especially thermal neutron flux, is conducted for restart of JMTR on FY2011. (author)

  17. Transparent Checkpoint-Restart for Hardware-Accelerated 3D Graphics

    OpenAIRE

    Nafchi, Samaneh Kazemi; Garg, Rohan; Cooperman, Gene

    2013-01-01

    Providing fault-tolerance for long-running GPU-intensive jobs requires application-specific solutions, and often involves saving the state of complex data structures spread among many graphics libraries. This work describes a mechanism for transparent GPU-independent checkpoint-restart of 3D graphics. The approach is based on a record-prune-replay paradigm: all OpenGL calls relevant to the graphics driver state are recorded; calls not relevant to the internal driver state as of the last graph...

  18. Safety Evaluation Report related to the restart of Davis-Besse Nuclear Power Station, Unit 1, following the event of June 9, 1985 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    On June 9, 1985, the Davis-Besse Nuclear Power Station, operated by the Toledo Edison Company, experienced a partial loss of main feedwater while the plant was at 90% power. The ensuing reactor trip was followed by spurious isolation of the steam geneators which initiated a chain of events involving a number of equipment malfunctions and several operator errors ultimately interrupting all feedwater for a short period of time. By the time operators were able to restore feedwater, both steam generators had dried out. A letter from the Director of the Office of Nuclear Reactor Regulation, pursuant to 10 CFR 50.54(f) of the Commission's regulations, confirmed that the Davis-Besse facility would not be restarted without NRC approval. The letter also requested that Toledo Edison submit its program for resolving numerous concerns identified by the staff. In response, the license submitted the Davis-Besse Course of Action report. The staff has reviewed that document and other supporting material submitted by the licensee; the staff's evaluation of that information is presented in this report

  19. Renewal of the WWR-K reactor operation: Collection of articles

    International Nuclear Information System (INIS)

    The book incorporates the papers about scientific researches carried out at the Almaty branch of the Institute of Atomic Energy of National Nuclear Centre of Republic of Kazakstan on substation of re-start of the WWR-K research reactor that was reformed in 1998, after 10 years of short-down. Collection consists of Introduction and 12 papers

  20. Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation)

    International Nuclear Information System (INIS)

    According to the demands for obtaining the licence for restarting the Ra reactor and the experimental operation this document includes the radiation monitoring measured data in the working space and environment of the RA reactor, i.e. Boris Kidric Institute. The meteorology measured data are included as well. All the measurements are performed according to the radiation protection program applied actually from the first reactor start-up at the end of 1959

  1. Annex V. Project re-start management experience Mochovce unit 3 and 4 NPP, Slovak Republic

    International Nuclear Information System (INIS)

    The purpose of this report is to describe the re-starting phase of DNPP MO 34 Project, mainly focusing on management issues. The report consists of three main parts dealing with the Project management organization during the three different periods of the of DNPP re-starting process. The first part (Chapter 2) begins with the general background information about the NPP and then focuses on the pre-feasibility and feasibility study. The second part (chapter 3) deals with the current status of the plant and it stresses the main steps of the DNPP Project resuming after completion of the feasibility study. The third part (Chapter 4) outlines the main issues related to the criticality of a typical DNPP Project (e.g. preservation and maintenance of equipments, updating to technological and regulatory requirements, human resources, protection of design documentation, etc.) and the measures implemented in MO 34. The hereby presented content reflects the experience and good practices concerning the following main management issues: - project control measures; - human resources; - preservation and maintenance of site installations, structures and equipment; - updating of the whole Project to meet licensing requirements and technology upgrades; - preservation of Project documentation. (author)

  2. Investigation of Loop Heat Pipe Survival and Restart After Extreme Cold Environment Exposure

    Science.gov (United States)

    Golliher, Eric; Ku, Jentung; Licari, Anthony; Sanzi, James

    2010-01-01

    NASA plans human exploration near the South Pole of the Moon, and other locations where the environment is extremely cold. This paper reports on the heat transfer performance of a loop heat pipe (LHP) exposed to extreme cold under the simulated reduced gravitational environment of the Moon. A common method of spacecraft thermal control is to use a LHP with ammonia working fluid. Typically, a small amount of heat is provided either by electrical heaters or by environmental design, such that the LHP condenser temperature never drops below the freezing point of ammonia. The concern is that a liquid-filled, frozen condenser would not restart, or that a thawing condenser would damage the tubing due to the expansion of ammonia upon thawing. This paper reports the results of an experimental investigation of a novel approach to avoid these problems. The LHP compensation chamber (CC) is conditioned such that all the ammonia liquid is removed from the condenser and the LHP is nonoperating. The condenser temperature is then reduced to below that of the ammonia freezing point. The LHP is then successfully restarted.

  3. The minimum cost optimal power flow problem solved via the restart homotopy continuation method

    International Nuclear Information System (INIS)

    The potential of the continuation method to solve the minimum cost optimal power (OPF) flow problem is assessed. Initially, the complete OPF problem is simplified by creating a sub-problem in which limits on functional variables are ignored. The restart homotopy continuation algorithm developed solves the sub-problem by manipulating the control variables to satisfy the optimality conditions of a family of relaxed sub-problems which converge to the desired solution. Particular features of this minimum cost problem are exploited to make the algorithm very efficient. If the solution to the sub-problem does not yield functional violations, then it becomes the solution to the complete OPF problem. However, if functional violations prevail, a new sub-problem is created by taking into consideration the functional violations, such that the newly created sub-problem maintains the same basic structure as its predecessor. This enables the restart homotopy continuation algorithm to be reapplied to solve the newly created sub-problem. The process is repeated until no functional violations prevail

  4. Rheology and FTIR studies of model waxy crude oils with relevance to gelled pipeline restart

    Energy Technology Data Exchange (ETDEWEB)

    Magda, J.J.; Guimeraes, K.; Deo, M.D. [Utah Univ., Salt Lake City, UT (United States). Dept. of Chemical Engineering; Venkatesan, R.; Montesi, A. [Chevron Energy Technology Co., Houston, TX (United States)

    2008-07-01

    Gels composed of wax crystals may sometimes form when crude oils are transported in pipelines when ambient temperatures are low. The gels may stop the pipe flow, making it difficult or even impossible to restart the flow without breaking the pipe. Rheology and FTIR techniques were used to study the problem and to characterize transparent model waxy crude oils in pipeline flow experiments. These model oils were formulated without any highly volatile components to enhance the reproducibility of the rheology tests. Results were presented for the time- and temperature-dependent rheology of the model waxy crude oils as obtained in linear oscillatory shear and in creep-recovery experiments. The model oils were shown to exhibit many of the rheological features reported for real crude oils, such as 3 distinct apparent yield stresses, notably static yield stress, dynamic yield stress, and elastic-limit yield stress. It was concluded that of the 3, the static yield stress value, particularly its time dependence, can best be used to predict the restart behaviour observed for the same gel in model pipelines.

  5. Data assimilation for unsaturated flow models with restart adaptive probabilistic collocation based Kalman filter

    Energy Technology Data Exchange (ETDEWEB)

    Man, Jun; Li, Weixuan; Zeng, Lingzao; Wu, Laosheng

    2016-06-01

    The ensemble Kalman filter (EnKF) has gained popularity in hydrological data assimilation problems. As a Monte Carlo based method, a relatively large ensemble size is usually required to guarantee the accuracy. As an alternative approach, the probabilistic collocation based Kalman filter (PCKF) employs the polynomial chaos to approximate the original system. In this way, the sampling error can be reduced. However, PCKF suffers from the so-called "curse of dimensionality". When the system nonlinearity is strong and number of parameters is large, PCKF could be even more computationally expensive than EnKF. Motivated by most recent developments in uncertainty quantification, we propose a restart adaptive probabilistic collocation based Kalman filter (RAPCKF) for data assimilation in unsaturated flow problems. During the implementation of RAPCKF, the important parameters are identified and active PCE basis functions are adaptively selected. The "restart" technology is used to eliminate the inconsistency between model parameters and states. The performance of RAPCKF is tested with numerical cases of unsaturated flow models. It is shown that RAPCKF is more efficient than EnKF with the same computational cost. Compared with the traditional PCKF, the RAPCKF is more applicable in strongly nonlinear and high dimensional problems.

  6. Closure theories with non-Gaussian restarts for truncated two-dimensional turbulence

    Science.gov (United States)

    Frederiksen, J. S.; Davies, A. G.; Bell, R. C.

    1994-09-01

    NonMarkovian closure theories, with and without non-Gaussian restarts, are compared with ensemble averaged direct numerical simulations (DNS) for severely truncated two-dimensional Navier-Stokes flows. Both the closures and DNS are formulated for discrete spectra relevant to flows on the doubly periodic domain allowing unambiguous comparisons between the closure and DNS results. We examine the performance of the direct interaction approximation (DIA), self-consistent field theory (SCFT) and local energy-transfer theory (LET) closures and are particularly interested in the reliability of cumulant update versions of these closures (CUDIA, CUSCFT, and CULET). In the latter, the potentially long time-history integrals are periodically truncated and the closures are restarted using a three-point cumulant as the new non-Gaussian initial conditions, thus yielding computationally much more efficient closures. In 80-day integrations, the DIA replicates the DNS results most faithfully in inviscid, viscous decay and forced dissipative experiments. With an update time of T=10 days, the CUDIA is particularly promising performing nearly as well but with some extra oscillations at intermediate times. The SCFT and particularly LET, have spurious oscillations in inviscid and viscous decay experiments; this is also the case, but to a greater degree, for the CUSCFT and CULET closures.

  7. RESTART simulation of non-Markov consecutive-k-out-of-n: F repairable systems

    Energy Technology Data Exchange (ETDEWEB)

    Villen-Altamirano, Jose, E-mail: jvillen@eui.upm.e [Departamento de Matematica Aplicada (E.U. Informatica), Universidad Politecnica de Madrid, Calle Arboleda s/n, 28031 Madrid (Spain)

    2010-03-15

    The reliability of consecutive-k-out-of-n: F repairable systems and (k-1)-step Markov dependence is studied. The model analyzed in this paper is more general than those of previous studies given that repair time and component lifetimes are random variables that follow a general distribution. The system has one repair service which adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly dependable systems, the RESTART method was used for the estimation of steady-state unavailability, MTBF and unreliability. Probabilities up to the order of 10{sup -16} have been accurately estimated with little computational effort. In this method, a number of simulation retrials are performed when the process enters regions of the state space where the chance of occurrence of a rare event (e.g., a system failure) is higher. The main difficulty for the application of this method is to find a suitable function, called the importance function, to define the regions. Given the simplicity involved in changing some model assumptions with RESTART, the importance function used in this paper could be useful for dependability estimation of many systems.

  8. RESTART simulation of non-Markov consecutive-k-out-of-n: F repairable systems

    International Nuclear Information System (INIS)

    The reliability of consecutive-k-out-of-n: F repairable systems and (k-1)-step Markov dependence is studied. The model analyzed in this paper is more general than those of previous studies given that repair time and component lifetimes are random variables that follow a general distribution. The system has one repair service which adopts a priority repair rule based on system failure risk. Since crude simulation has proved to be inefficient for highly dependable systems, the RESTART method was used for the estimation of steady-state unavailability, MTBF and unreliability. Probabilities up to the order of 10-16 have been accurately estimated with little computational effort. In this method, a number of simulation retrials are performed when the process enters regions of the state space where the chance of occurrence of a rare event (e.g., a system failure) is higher. The main difficulty for the application of this method is to find a suitable function, called the importance function, to define the regions. Given the simplicity involved in changing some model assumptions with RESTART, the importance function used in this paper could be useful for dependability estimation of many systems.

  9. Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

    International Nuclear Information System (INIS)

    An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

  10. Fast breeder reactor. The past, the present and the future. (7) History of fast reactor development in Japan - 2

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 2, this seventh lecture presented the development of the prototype FBR (MONJU) and design studies of the demonstration reactor. The MONJU started operation in 1994, but a sodium leakage in its secondary heat transfer system occurred during performance tests in 1995. It has not operated since and activities for restart are conducted. Since 1997 design studies of the demonstration FBR have been conducted to reflect the MONJU sodium leakage accident and also establish its economic competitiveness with advanced LWR. (T. Tanaka)

  11. A review of fast reactor activities in Italy

    International Nuclear Information System (INIS)

    In this paper, Italian activities on liquid metal fast reactors are shown for the period May 1995 - April 1996. During this period the ENEA collaboration with General Electric on ALMR came to an end as a consequence of the reduced effort on the design development. Nevertheless ENEA completed the studies on the PRISM Mod B oxide burner core, the neutronic configuration of which was presented at last year's meeting. Some results of the dynamic calculations are shown. ENEA participated in the IAEA/EC benchmark on the comparative calculations for severe accident in BN-800 reactor. A complete neutronic evaluation has been made including power distribution, Doppler, sodium void and material coefficients. Activities on seismic isolation are also outlined. The Italian contribution to SPX restart and operation is described; some information about the complementary convention of the Nersa society is given, together with the Italian industry participation in the SPX restart. (author)

  12. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  13. Life stage differences in resident coping with restart of the Three Mile Island nuclear generating facility

    International Nuclear Information System (INIS)

    A study of residents who remained in the vicinity of Three Mile Island (TMI) immediately following the restart of the nuclear generating plant revealed that older residents employed a more emotion-focused coping style in the face of this event than did younger residents. Coping style was, however, unrelated to the level of psychological symptoms for these older residents, whereas demographic variables were related. Among younger residents, on the other hand, coping style was related to the level of psychological symptoms, whereas demographic variables were not. Among younger residents, emotion-focused coping was associated with more symptoms and problem-focused coping was associated with fewer symptoms, contradicting previous findings among TMI area residents

  14. Asynchronous Checkpoint Migration with MRNet in the Scalable Checkpoint / Restart Library

    Energy Technology Data Exchange (ETDEWEB)

    Mohror, K; Moody, A; de Supinski, B R

    2012-03-20

    Applications running on today's supercomputers tolerate failures by periodically saving their state in checkpoint files on stable storage, such as a parallel file system. Although this approach is simple, the overhead of writing the checkpoints can be prohibitive, especially for large-scale jobs. In this paper, we present initial results of an enhancement to our Scalable Checkpoint/Restart Library (SCR). We employ MRNet, a tree-based overlay network library, to transfer checkpoints from the compute nodes to the parallel file system asynchronously. This enhancement increases application efficiency by removing the need for an application to block while checkpoints are transferred to the parallel file system. We show that the integration of SCR with MRNet can reduce the time spent in I/O operations by as much as 15x. However, our experiments exposed new scalability issues with our initial implementation. We discuss the sources of the scalability problems and our plans to address them.

  15. Implementation plan for the Waste Experimental Reduction Facility Restart Operational Readiness Review

    International Nuclear Information System (INIS)

    The primary technical objective for the WERF Restart Project is to assess, upgrade where necessary, and implement management, documentation, safety, and operation control systems that enable the resumption and continued operation of waste treatment and storage operations in a manner that is compliant with all environment, safety, and quality requirements of the US Department of Energy and Federal and State regulatory agencies. Specific processes that will be resumed at WERF include compaction of low-level compatible waste; size reduction of LLW, metallic and wood waste; incineration of combustible LLW and MLLW; and solidification of low-level and mixed low-level incinerator bottom ash, baghouse fly ash, and compatible sludges and debris. WERF will also provide for the operation of the WWSB which includes storage of MLLW in accordance with Resource Conservation and Recovery Act requirements

  16. The crystal structure of Neisseria gonorrhoeae PriB reveals mechanistic differences among bacterial DNA replication restart pathways

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Jinlan; George, Nicholas P.; Duckett, Katrina L.; DeBeer, Madeleine A.P.; Lopper, Matthew E. (UDRI); (UW-MED)

    2010-05-25

    Reactivation of repaired DNA replication forks is essential for complete duplication of bacterial genomes. However, not all bacteria encode homologs of the well-studied Escherichia coli DNA replication restart primosome proteins, suggesting that there might be distinct mechanistic differences among DNA replication restart pathways in diverse bacteria. Since reactivation of repaired DNA replication forks requires coordinated DNA and protein binding by DNA replication restart primosome proteins, we determined the crystal structure of Neisseria gonorrhoeae PriB at 2.7 {angstrom} resolution and investigated its ability to physically interact with DNA and PriA helicase. Comparison of the crystal structures of PriB from N. gonorrhoeae and E. coli reveals a well-conserved homodimeric structure consisting of two oligosaccharide/oligonucleotide-binding (OB) folds. In spite of their overall structural similarity, there is significant species variation in the type and distribution of surface amino acid residues. This correlates with striking differences in the affinity with which each PriB homolog binds single-stranded DNA and PriA helicase. These results provide evidence that mechanisms of DNA replication restart are not identical across diverse species and that these pathways have likely become specialized to meet the needs of individual organisms.

  17. 78 FR 34639 - Supplemental Final Environmental Impact Statement for the Restart of Healy Power Plant Unit #2

    Science.gov (United States)

    2013-06-10

    ...) since 1967. Healy Unit 2 is a 50 MW coal-fired steam generator owned by AIDEA, which underwent test operation for two years as part of DOE's project. Unit 2 has been in warm layup since late 1999. The SFEIS updates information in DOE's FEIS and considers impacts of restarting Unit 2 for commercial operation....

  18. Getting to First Flight: Equipping Space Engineers to Break the Start-Stop-Restart Cycle

    Science.gov (United States)

    Singer, Christopher E.; Dumbacher, Daniel L.

    2010-01-01

    The National Aeronautics and Space Administration s (NASA s) history is built on a foundation of can-do strength, while pointing to the Saturn/Apollo Moon missions in the 1960s and 1970s as its apex a sentiment that often overshadows the potential that lies ahead. The chronicle of America s civil space agenda is scattered with programs that got off to good starts with adequate resources and vocal political support but that never made it past a certain milestone review, General Accountability Office report, or Congressional budget appropriation. Over the decades since the fielding of the Space Shuttle in the early 1980s, a start-stop-restart cycle has intervened due to many forces. Despite this impediment, the workforce has delivered engineering feats such as the International Space Station and numerous Shuttle and science missions, which reflect a trend in the early days of the Exploration Age that called for massive infrastructure and matching capital allocations. In the new millennium, the aerospace industry must respond to transforming economic climates, the public will, national agendas, and international possibilities relative to scientific exploration beyond Earth s orbit. Two pressing issues - workforce transition and mission success - are intertwined. As this paper will address, U.S. aerospace must confront related workforce development and industrial base issues head on to take space exploration to the next level. This paper also will formulate specific strategies to equip space engineers to move beyond the seemingly constant start-stop-restart mentality to plan and execute flight projects that actually fly.

  19. Structure and Function of the PriC DNA Replication Restart Protein.

    Science.gov (United States)

    Wessel, Sarah R; Cornilescu, Claudia C; Cornilescu, Gabriel; Metz, Alice; Leroux, Maxime; Hu, Kaifeng; Sandler, Steven J; Markley, John L; Keck, James L

    2016-08-26

    Collisions between DNA replication complexes (replisomes) and barriers such as damaged DNA or tightly bound protein complexes can dissociate replisomes from chromosomes prematurely. Replisomes must be reloaded under these circumstances to avoid incomplete replication and cell death. Bacteria have evolved multiple pathways that initiate DNA replication restart by recognizing and remodeling abandoned replication forks and reloading the replicative helicase. In vitro, the simplest of these pathways is mediated by the single-domain PriC protein, which, along with the DnaC helicase loader, can load the DnaB replicative helicase onto DNA bound by the single-stranded DNA (ssDNA)-binding protein (SSB). Previous biochemical studies have identified PriC residues that mediate interactions with ssDNA and SSB. However, the mechanisms by which PriC drives DNA replication restart have remained poorly defined due to the limited structural information available for PriC. Here, we report the NMR structure of full-length PriC from Cronobacter sakazakii PriC forms a compact bundle of α-helices that brings together residues involved in ssDNA and SSB binding at adjacent sites on the protein surface. Disruption of these interaction sites and of other conserved residues leads to decreased DnaB helicase loading onto SSB-bound DNA. We also demonstrate that PriC can directly interact with DnaB and the DnaB·DnaC complex. These data lead to a model in which PriC acts as a scaffold for recruiting DnaB·DnaC to SSB/ssDNA sites present at stalled replication forks. PMID:27382050

  20. Safety Evaluation Report related to the restart of Rancho Seco Nuclear Generating Station, Unit 1, following the event of December 26, 1985 (Docket No. 50-312)

    International Nuclear Information System (INIS)

    On December 26, 1985, the Rancho Seco Nuclear Generating Station, owned and operated by the Sacramento Municipal Utility District (SMUD), experienced a loss of dc power within the integrated control system (ICS) while the plant was at 76% power. The ensuing reactor trip was followed by a rapid overcooling transient and automatic initiation of the safety features actuation system (SFAS). The overcooling transient continued until ICS dc power was restored 26 minutes after its loss. Two letters from the NRC Region V Administrator (dated December 26, 1985) confirmed that the Rancho Seco plant would not be returned to power operation until SMUD (the licensee) had provided the NRC with an assessment of the root cause of the transient and a justification as to why the Rancho Seco facility is ready to resume power operation. In response, the licensee submitted the ''Rancho Seco Action Plan for Performance Improvement'' on July 3, 1986; revisions to that action plan were submitted on December 15, 1986 and February 28, 1987. The NRC staff has reviewed the action plan and numerous other supporting documents submitted by the licensee. The staff's evaluation of the information supporting restart of Rancho Seco is presented in this safety evaluation report

  1. L-Reactor Habitat Mitigation Study

    International Nuclear Information System (INIS)

    The L-Reactor Fish and Wildlife Resource Mitigation Study was conducted to quantify the effects on habitat of the L-Reactor restart and to identify the appropriate mitigation for these impacts. The completed project evaluated in this study includes construction of a 1000 acre reactor cooling reservoir formed by damming Steel Creek. Habitat impacts identified include a loss of approximately 3,700 average annual habitat units. This report presents a mitigation plan, Plan A, to offset these habitat losses. Plan A will offset losses for all species studied, except whitetailed deer. The South Carolina Wildlife and Marine Resources Department strongly recommends creation of a game management area to provide realistic mitigation for loss of deer habitats. 10 refs., 5 figs., 3 tabs

  2. Evaluation of the Steel Creek ecosystem in relation to the propsed restart of the L-reactor: interim report

    International Nuclear Information System (INIS)

    The ecological assessment of the Steel Creek ecosystempresents information concerning the fall and winter utilization of Steel Creek habitats. An understanding of the wintering behavior of these species is necessary to assess the impact of thermal effluents on these species. Waterfowl use of the Steel Creek habitats was extensive during winter 1981-1982. The habitat was used for both foraging and roosting. Mallards and wood ducks were the most important local game species. Observations of the thermal deltas indicated substantial migratory waterfowl use of areas near Four Mile Creek, while no ducks were observed in the vicinity of Pen Branch. Studies of the American alligator through the winter of 1981-1982 indicated that this species remains active through the winter. It has also been established that the Florida mud turtle, occurs in Steel Creek. Fish collections through the fall of 1981 and winter of 1981-1982 yielded 49 species. Considerable differences in overall fish species diversity and changes in fish species composition were observed. Substantial variations in radiocesium concentrations exist among species although all species were exposed to the same concentrations of radiocesium in water. Substantially less radiocesium remains in the system than was anticipated. Substantial differences occur in different stream segments as well as the vertical distribution of the isotope through the soil profile. The studies to date document that animal use of the Steel Creek cosystems is substantial

  3. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) in India is ready for restart. Satisfactory progress has been made in the design of Prototype Fast Breeder Reactor (PFBR). Conceptual design work for the important systems and components has been completed. Cost estimation is in progress. Detailed project report for the financial sanction is under completion stage and is planned to be submitted to the Government this year. Draft Safety criteria prepared by a sub-committee on behalf of the Regulatory Board have been discussed and will be issued shortly. (author)

  4. The BR2-material testing reactor and its major contribution to the reactor material, fuel and safety research

    International Nuclear Information System (INIS)

    The BR2 was shutdown at the end of June 1995 for a programme of extensive refurbishment after more than 30 years utilization. The beryllium matrix was replaced and the aluminum vessel inspected and requalified for the envisaged 15 years life extension. Other aspects of the refurbishment programme were aimed at reliability and availability of the installations, safety of operation and compliance with modem safety standards. The reactor was restarted in April 1997. This paper deals with aspects of this refurbishment in general as well as the ongoing experimental projects in the areas of reactor material, fuel behaviour and safety research. (author)

  5. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  6. A study on people's awareness about the restarting and decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    In this study, we conducted two questionnaire surveys targeting a total of 918 respondents living in the cities of Kyoto, Osaka and Kobe, in order to elucidate people's awareness of three things: 1) restart of nuclear power plants; 2) extension of the operation period of aging plants; and 3) decommissioning. The results are as follows: 1) People who think that electrical power companies voluntarily take higher safety measures trust the power companies and do not oppose the restart of the nuclear power plants, as compared to people who think that power companies only meet the requirements set by the nuclear regulatory agency. 2) When people were given information about aging measures and conforming to new regulatory standards, their anxiety toward the operation of aging plants was reduced. 3) People thought that decommissioning work was important for society. However, a small number of people thought it was a job worthwhile doing. (author)

  7. Global Convergence of a New restarting Conjugate Gradient Method for Nonlinear Optimizations

    Institute of Scientific and Technical Information of China (English)

    SUNQing-ying

    2003-01-01

    Conjugate gradient optimization algorithms depend on the search directions.with different choices for the parameters in the search directions.In this note,by combining the nice numerical performance of PR and HS methods with the global convergence property of the class of conjugate gradient methods presented by HU and STOREY(1991),a class of new restarting conjugate gradient methods is presented.Global convergences of the new method with two kinds of common line searches,are proved .Firstly,it is shown that,using reverse modulus of continuity funciton and forcing function,the new method for solving unconstrained optimization can work for a continously differentiable function with Curry-Altman's step size rule and a bounded level set .Secondly,by using comparing technique,some general convergence propecties of the new method with other kind of step size rule are established,Numerical experiments show that the new method is efficient by comparing with FR conjugate gradient method.

  8. Control assembly for controlling a fuel cell system during shutdown and restart

    Science.gov (United States)

    Venkataraman, Ramki; Berntsen, George; Carlson, Glenn L.; Farooque, Mohammad; Beachy, Dan; Peterhans, Stefan; Bischoff, Manfred

    2010-06-15

    A fuel cell system and method in which the fuel cell system receives and an input oxidant gas and an input fuel gas, and in which a fuel processing assembly is provided and is adapted to at least humidify the input fuel gas which is to be supplied to the anode of the fuel cell of the system whose cathode receives the oxidant input gas via an anode oxidizing assembly which is adapted to couple the output of the anode of the fuel cell to the inlet of the cathode of the fuel cell during normal operation, shutdown and restart of the fuel cell system, and in which a control assembly is further provided and is adapted to respond to shutdown of the fuel cell system during which input fuel gas and input oxidant gas cease to be received by the fuel cell system, the control assembly being further adapted to, when the fuel cell system is shut down: control the fuel cell system so as to enable a purging gas to be able to flow through the fuel processing assembly to remove humidified fuel gas from the processing assembly and to enable a purging gas to be able to flow through the anode of the fuel cell.

  9. Efficient two-dimensional magnetotellurics modelling using implicitly restarted Lanczos method

    Indian Academy of Sciences (India)

    Krishna Kumar; Pravin K Gupta; Sri Niwas

    2011-08-01

    This paper presents an efficient algorithm, FDA2DMT (Free Decay Analysis for 2D Magnetotellurics (MT)), based on eigenmode approach to solve the relevant partial differential equation, for forward computation of two-dimensional (2D) responses. The main advantage of this approach lies in the fact that only a small subset of eigenvalues and corresponding eigenvectors are required for satisfactory results. This small subset (pre-specified number) of eigenmodes are obtained using shift and invert implementation of Implicitly Restarted Lanczos Method (IRLM). It has been established by experimentation that only 15–20% smallest eigenvalue and corresponding eigenvectors are sufficient to secure the acceptable accuracy. Once the single frequency response is computed using eigenmode approach, the responses for subsequent frequencies can be obtained in negligible time. Experiment design results for validation of FDA2DMT are presented by considering two synthetic models from COMMEMI report, Brewitt-Taylor and Weaver (1976) model and a field data based model from Garhwal Himalaya.

  10. Critical restarting time window wax concentration effect for reducing waxy crude transportation risk

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, Jamal M; Salleh, Intan Khalida [PETRONAS Research (Malaysia)], email: Jamal.Ibrahim@pet.hw.co.uk; Toma, Peter [P.R.Toma Consulting Ltd (Canada)], email: ptoma@telus.net

    2010-07-01

    Critical restarting Time Window (CRTW) is a field specification that indicates the maximum value of the shut-in time for normal resumption of steady flow conditions. This paper studies field wax composition, using offshore temperatures and pipe diameters to demonstrate the salient effect of wax composition-WAT on CRTW. In fact, the CRTW and the wax appearance temperature (WAT) are two competing factors controlling the transient heat transfer during shut-in periods. Thus, proper assessment of WAT and radial-time-temperature curves during transient cool-down are both major pieces of essential information required to carrying out this research. In order to release the effect of crystallization heat a finite element heat transfer model (FEHT) was used. The transient heat model designed for a specific pipe geometry was also used to calculate radius-specific time-temperature functions. Finally, the conclusions and results of this research are an asset in understanding a complex of mitigation technologies used to reduce the CRTW for offshore production operations.

  11. Ghost Process: a Sound Basis to Implement Process Duplication, Migration and Checkpoint-Restart in Linux Clusters

    OpenAIRE

    Vallée, Geoffroy; Lottiaux, Renaud; Margery, David; Morin, Christine; Berthou, Jean-Yves

    2005-01-01

    Process management mechanisms (process duplication, migration and checkpoint-restart) are very useful for high performance and high availability in clustering systems. The single system image approach aims at providing a global process management service with mechanisms for process checkpoint, process migration and process duplication. In this context, a common mechanism for process virtualization is highly desirable but traditional operating systems do not provide such a mechanism. This pape...

  12. Knowledge management implementation on the restart of the nuclear power plant construction

    International Nuclear Information System (INIS)

    Full text: Restarting the NPP construction after 10 years of inactivity is a process that involves many preliminary tasks associated with the transition period to get the project started again. Implementing a KM program during the preparatory phase motivates the personnel and facilitates the completion of these additional activities. Human Resources Motivation: Manpower is the most critical aspect to consider at the moment of restarting the NPP project. The reduced engineering teams left at the NPP lost their motivation as a result of the absence of project requirements. These groups, which were responsible of key activities in the past, and now assigned to other tasks, must be reinserted to the schedule and functions required by the project management. Moreover, they constitute the core that would transfer knowledge to the future personnel. Therefore, it is a good practice to include these engineering groups from the very beginning of the KM development. It is proved that the participation of these groups in the KM design and definition, in the knowledge map building, in identifying the domains and performing critical knowledge analysis by means of workshops, and in meetings and individual interviews facilitates the reactivation of them. The demands from the Knowledge Management Project create a good atmosphere to stimulate sharing and competences development. Capturing Experts' Knowledge. During the years of inactivity of the plant construction many professionals and specialists that belonged to the original project teams left the organization taking with them their data and information related to the project evolution, and valuable undocumented knowledge. Documented meetings between current and past experts, or through an Experts Consulting Group articulates this tacit knowledge, and provides a source of answers about previous situations, taken decisions and critical issues. Furthermore, implementing a feedback program prevents the risk of knowledge loss due to

  13. Annual report of the Neutron Irradiation and Testing Reactor Center. FY 2007. April 1, 2007 - March 31, 2008

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR), achieving the first criticality in March 1968, has been used to test the durability and integrity of reactor fuels and components, basic nuclear research, production of radioisotopes (RIs), and other purposes. The JMTR, however, was halted in August 2006 after its 165th cycle operation, and is currently undergoing partial renewal of the apparatus and installation of new irradiation equipment, aiming at restarting from 2011. In addition, to cope with strong requests from users to improve the usability of the JMTR, efforts are being made to increase reactor operating efficiency, shorten the turnaround time for obtaining results, and conduct other necessary tasks for the JMTR to recommence reoperation. The present report summarizes the activities carried out in 2007 for the refurbishment and restart of the JMTR. (author)

  14. R- AND P- REACTOR VESSEL IN-SITU DECOMISSIONING VISUALIZATION

    Energy Technology Data Exchange (ETDEWEB)

    Vrettos, N.; Bobbitt, J.; Howard, M.

    2010-06-07

    The R- & P- Reactor facilities were constructed in the early 1950's in response to Cold War efforts. The mission of the facilities was to produce materials for use in the nation's nuclear weapons stockpile. R-Reactor was removed from service in 1964 when President Johnson announced a slowdown of he nuclear arms race. PReactor continued operation until 1988 until the facility was taken off-line to modernize the facility with new safeguards. Efforts to restart the reactor ended in 1990 at the end of the Cold War. Both facilities have sat idle since their closure and have been identified as the first two reactors for closure at SRS.

  15. Flowsheet for shear/leach processing of N Reactor fuel at PUREX

    Energy Technology Data Exchange (ETDEWEB)

    Enghusen, M.B.

    1995-04-13

    This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant.

  16. Flowsheet for shear/leach processing of N Reactor fuel at PUREX

    International Nuclear Information System (INIS)

    This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant

  17. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues

  18. A Preliminary Analysis of Cometary Dust in the 1st Year of the NEOWISE Restarted Mission

    Science.gov (United States)

    Kramer, Emily A.; Bauer, James M.; Fernández, Yanga R.; Mainzer, Amy K.; Grav, Tommy; Masiero, Joseph R.; Nugent, Carolyn R.; Sonnett, Sarah; Cutri, Roc; Stevenson, Rachel

    2015-11-01

    As some of the most pristine objects in the Solar System, comets present an opportunity to understand the mechanics and chemistry of the planetary formation era. By studying a large number of comets in different dynamical classes, we can better understand the ensemble properties of the different classes, and begin to characterize the evolution that may have occurred since their formation.In late 2013, the WISE spacecraft was brought out of hibernation, and renamed NEOWISE with a renewed goal to detect and characterize small bodies using its 3.4 and 4.6-micron bands. Survey operations began in December 2013 [1], and the first year of data was publicly released in March 2015 [2]. During the course of the first year of the restarted mission, over 60 comets were serendipitously detected by NEOWISE at heliocentric distances between ~1-7.5 AU, including 3 newly discovered comets. The comets detected were split roughly evenly between short-period and long-period comets, and many displayed extended dust structures. Several of the comets were detected multiple times over the course of the year, and some were also seen during the prime WISE mission. This long baseline allows for an intriguing analysis of long-term cometary behavior.NEOWISE has sampled the behavior of these comet dynamical sub-types over the thermal infrared and near-infrared reflected-light regimes, where effects from different particle size ranges of dust may dominate the morphologies and observed fluxes. We present a preliminary analysis of the cometary dust seen in these data, including dynamical models to constrain the sizes and ages of the dust particles. We discuss how these results compare to those obtained for the comets seen in the 12 and 22-micron WISE prime mission data.Acknowledgments: This publication makes use of data products from (1) WISE, which is a joint project of UCLA and JPL/Caltech, funded by NASA; and (2) NEOWISE, which is a project of JPL/Caltech, funded by the Planetary Science

  19. Direct versus Indirect Treatment for Preschool Children who Stutter: The RESTART Randomized Trial.

    Directory of Open Access Journals (Sweden)

    Caroline de Sonneville-Koedoot

    Full Text Available Stuttering is a common childhood disorder. There is limited high quality evidence regarding options for best treatment. The aim of the study was to compare the effectiveness of direct treatment with indirect treatment in preschool children who stutter.In this multicenter randomized controlled trial with an 18 month follow-up, preschool children who stutter who were referred for treatment were randomized to direct treatment (Lidcombe Program; n = 99 or indirect treatment (RESTART-DCM treatment; n = 100. Main inclusion criteria were age 3-6 years, ≥3% syllables stuttered (%SS, and time since onset ≥6 months. The primary outcome was the percentage of non-stuttering children at 18 months. Secondary outcomes included stuttering frequency (%SS, stuttering severity ratings by the parents and therapist, severity rating by the child, health-related quality of life, emotional and behavioral problems, and speech attitude.Percentage of non-stuttering children for direct treatment was 76.5% (65/85 versus 71.4% (65/91 for indirect treatment (Odds Ratio (OR, 0.6; 95% CI, 0.1-2.4, p = .42. At 3 months, children treated by direct treatment showed a greater decline in %SS (significant interaction time x therapy: β = -1.89; t(282.82 = -2.807, p = .005. At 18 months, stuttering frequency was 1.2% (SD 2.1 for direct treatment and 1.5% (SD 2.1 for indirect treatment. Direct treatment had slightly better scores on most other secondary outcome measures, but no differences between treatment approaches were significant.Direct treatment decreased stuttering more quickly during the first three months of treatment. At 18 months, however, clinical outcomes for direct and indirect treatment were comparable. These results imply that at 18 months post treatment onset, both treatments are roughly equal in treating developmental stuttering in ways that surpass expectations of natural recovery. Follow-up data are needed to confirm these findings in the longer term

  20. N Reactor

    Data.gov (United States)

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

  1. The Last Twenty Years of Experience with Fast Reactors in Japan

    International Nuclear Information System (INIS)

    Fast reactor development experience gained in Japan in the last twenty years is summarized in this paper. In this twenty years, the safety, reliability and economic goals of fast reactors have become more ambitious than in the past. However, twenty years of progress have shown that the domestic commercialized sodium cooled fast reactor (SFR) concept, the Japanese SFR, could achieve those targets discussed in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) projects. The Monju prototype fast breeder reactor is finally going to restart by the end of this Japanese fiscal year (March 2010) and will take on the role of a technology and human resource development centre from both a domestic and an international point of view. (author)

  2. Analysis of factors associated with hesitation to restart farming after depopulation of animals due to 2010 foot-and-mouth disease epidemic in Japan.

    Science.gov (United States)

    Kadowaki, Hazumu; Kayano, Taishi; Tobinaga, Takaharu; Tsutsumi, Atsuro; Watari, Michiko; Makita, Kohei

    2016-09-01

    An outbreak of foot-and-mouth disease (FMD) occurred in Miyazaki Prefecture, Japan, in 2010. This epidemic was controlled with culling and vaccination, and resulted in the death of nearly 290,000 animals. This paper describes the factors associated with hesitation to restart farming after the epidemic. A questionnaire survey was conducted to assess the mental health of farmers one year after the end of the FMD epidemic in affected areas, and univariate and multivariable analyses were performed. Of 773 farms which had answered the question about restart farming, 55.4% (428/773) had resumed or were planning to resume operation. The farms hesitated restarting were characterized by small scale (P=0.06) and having multiple sources of income (Plabor force, had an alternative business or were mentally distressed during disease control. PMID:27149890

  3. Westinghouse independent safety review of Savannah River production reactors

    International Nuclear Information System (INIS)

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

  4. Westinghouse independent safety review of Savannah River production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

    1989-04-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

  5. Servicing and safe operation constraints for initial and compact loading of WWR-K research reactor core

    International Nuclear Information System (INIS)

    The WWR-K research reactor with the design power of 10 MWt was put into operation in 1967 year. Its operation was ceased in 1988 year due to the absence of its safety validation under seismicity conditions (9 points by MSK-64 scale). After the activity aimed at enhancement of the reactor and technological equipment seismic resistance was completed in 1997 year, a decision to restart the reactor was made. The enhancement activity also implied a revision of the reactor core configuration and its optimization to reduce the environmental impact in case of unexpected events.There was conduction of analysis of parameters of an initial compact loading of WWR-K reactor and establishment of its reactivity margin rated for the beginning of the operating cycle. The margin allows a continuous operation of the reactor during 10 days. The nominal reactor power value is 6 MWt

  6. Reactor Physics

    International Nuclear Information System (INIS)

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

  7. Reactor Physics

    International Nuclear Information System (INIS)

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

  8. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  9. A proposal on restart rule of nuclear power plants with piping having local wall thinning subjected to an earthquake. Former part. Aiming at further application

    International Nuclear Information System (INIS)

    Restart rule of nuclear power plants (NPPs) with piping having local wall thinning subjected to an earthquake was proposed taking account of local wall thinning, seismic effects and restart of NPPs with applicability of 'Guidelines for NPP Response to an Earthquake (EPRI NP-6695)' in Japan. Japan Earthquake Damage Intensity Scale (JEDIS) and Earthquake Ground Motion Level (EGML) were introduced. JEDIS was classified into four scales obtained from damage level of components and structures of NPPs subjected to an earthquake, while EGML was divided into four levels by safe shutdown earthquake ground motion (So), elastic design earthquake ground motion (Sd) and design earthquake ground motion (Ss). Combination of JEDIS and EGML formulated 4 x 4 matrix and determined detailed conditions of restart of NPPs. As a response to an earthquake, operator walk inspections and evaluation of earthquake ground motion were conducted to know the level of JEDIS. JEDIS level requested respective allowable conditions of restart of NPP, which were scale level dependent and consisted of weighted combination of damage inspection (operator walk inspections, focused inspections/tests and expanded inspections), integrity evaluation and repair/replacement. If JEDIS were assigned greater than 3 with expanded inspections, inspection of piping with local wall thinning, its integrity evaluation and repair/replacement if necessary were requested. Inspection and evaluation of piping with local wall thinning was performed based on JSME or ASME codes. Detailed work flow charts were presented. Carbon steel piping and elbow was chosen for evaluation. (T. Tanaka)

  10. 40 CFR 60.1635 - What must I do if I close my municipal waste combustion unit and then restart my municipal waste...

    Science.gov (United States)

    2010-07-01

    ..., you must complete emission control retrofit and meet the emission limits and good combustion practices... waste combustion unit and then restart my municipal waste combustion unit? 60.1635 Section 60.1635... Combustion Units Constructed on or Before August 30, 1999 Model Rule-Increments of Progress § 60.1635...

  11. Identification of the glycerol kinase gene and its role in diapause embryo restart and early embryo development of Artemia sinica.

    Science.gov (United States)

    Cheng, Cheng; Yao, Feng; Chu, Bing; Li, Xuejie; Liu, Yan; Wu, Yang; Mei, Yanli; Wang, Peisheng; Hou, Lin; Zou, Xiangyang

    2014-03-01

    Glycerol kinase (GK) catalyzes the rate-limiting step in glycerol utilization by transferring a phosphate from ATP to glycerol, yielding glycerol 3-phosphate, which is an important intermediate for both energy metabolism and glycerolipid production. Artemia sinica has an unusual diapause process under stress conditions of high salinity, low temperature and lack of food. In the process, diapause embryos of A. sinica (brine shrimp) accumulate high concentrations of glycerol as a cryoprotectant to prevent low temperature damage to embryos. Upon embryo restart, glycerol is converted into glucose and other carbohydrates. Therefore, GK plays an important role in the diapause embryo restart process. However, the role of GK in diapause termination of embryo development in A. sinica remains unknown. In the present study, a 2096 bp full-length cDNA of gk from A. sinica (As-gk) was obtained, encoding putative 551 amino acids, 60.6 kDa protein. As a crucial enzyme in glycerol uptake and metabolism, GK has been conserved structurally and functionally during evolution. The expression pattern of As-gk was investigated by quantitative real-time PCR and Western blotting. Expression locations of As-gk were analyzed using in situ hybridization. As-gk was widely distributed in the early embryo and several main parts of Artemia after differentiation. The expression of As-GK was also induced by stresses such as cold exposure and high salinity. This initial research into the expression pattern and stress response of GK in Artemia provides a sound basis for further understanding of the function and regulation of genes in early embryonic development in A. sinica and the stress response. PMID:24365596

  12. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  13. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

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

  14. Conversion at the WWR-K-Almaty research reactor

    International Nuclear Information System (INIS)

    In the former USSR the research reactor WWR-K (Kazakhstan National Nuclear Centre) for many years was used, mainly, for investigations ordered by Ministry of Defense. During last years crucial re-orientation to peaceful goals has been overtaken. However due to poor budget the conversion process evolves extremely slowly. Due to the Japanese Government financing the ISTC (International Scientific Technical Center) Project K-012, restart of the reactor WWR-K after its 9 years shutdown becomes possible. We have high-qualified specialists and powerful reactor facility, and we wouldn't want to lose both. For surviving reactor operation we propose the studies which can be interesting for foreign investors. The reactor WWR-K is located in high-seismicity zone. The investigations of safety items related to this fact can be interesting, for instance, for Japan. Following studies are proposed. 1) Creation of the vault for dry storage of spent fuel at the conditions of high regional seismicity. 2) Provision of the required extent of safety for the reactor burial of radioactive waste at the conditions of high regional seismicity. 3) Creation of the automate system of permanent seismic monitoring for the reactor WWR-K located in the high-seismicity region

  15. 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.

  16. GOLEM: a versatile computer code for reactor neutronic calculation advances in qualification of the different modules

    International Nuclear Information System (INIS)

    The last 12 years studies about the CABRI, SCARABEE and PHEBUS projects are summarized. It describes the object and the genesis of the cores, the evolution of the core concept and the associated neutronic problems. The calculational scheme used is presented, together with its qualification. The formalism, and the qualification of the different modules of GOLEM are presented. COXYS: module of physical analysis in order to determine the best energetic and spatial mesh for the case of interest. GOLU.B: input data management module. VAREC: calculation module of perturbations due to materials enables to compute perturbed flux and reactivity variation. VARYX: calculation module of geometric perturbations. TRACASYN: module of 3D power shape calculation. Finally TRACASTORE: module of management and graphic exploitation of results. Then, one gives utilization directions for these different modules. Qualification results show that GOLEM is able to analyse the fine physics of many various cases, to calculate by perturbation effects greater than 5000 pcm, to rebuild perturbed flux with margins near 3% for difficult situations, like reactor voiding or spectral or spectral variation in a PWR. Furthermore, 3D hot spots are calculated within margins of a magnitude comparable to experimental ones

  17. Method for loading, operating, and unloading a ball-bed nuclear reactor

    International Nuclear Information System (INIS)

    This patent describes a method of operating a ball-bed nuclear reactor with fuel element balls. Some have a fissionable material content different from that of others of the balls. It consists of: initially partly filling a reactor core with fuel balls of sufficient fissionable material content for establishing criticality and a desired level of power production at the completion of the partial filling and then, without any further filling of the reactor cavern, starting reactor operation; thereafter without any removal of fuel balls from the reactor cavern, filling fuel balls continually or in groups at relatively short intervals into the reactor cavern during increasing burning up of the fuel balls already, for compensation of the diminishing fissionable material content of the reactor core constituted by the fuel balls until a final total quantity of filling is reached; after the final filling quantity is reached and burning up has occurred, shutting down the reactor, cooling it off, releasing the pressure in the cavern, and thereafter unloading all the fuel balls from the reactor cavern, unloading being begun when the reactor is shut down and being completed before the reactor is restarted

  18. 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)

  19. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

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

  20. 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

  1. Reactor container

    International Nuclear Information System (INIS)

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

  2. Modeling of the re-starting of waxy crude oil flows in pipelines; Modelisation du redemarrage des ecoulements de bruts paraffiniques dans les conduites petrolieres

    Energy Technology Data Exchange (ETDEWEB)

    Vinay, G.

    2005-11-15

    Pipelining crude oils that contain large proportions of paraffins can cause many specific difficulties. These oils, known as waxy crude oils, usually exhibit high 'pour point', where this temperature is higher than the external temperature conditions surrounding the pipeline. During the shutdown, since the temperature decreases in the pipeline, the gel-like structure builds up and the main difficulty concerns the issue of restarting. This PhD attempts to improve waxy crude oil behaviour understanding thanks to experiment, modelling and numerical simulation in order to predict more accurately time and pressure required to restart the flow. Using various contributions to the literature, waxy crude oils are described as viscoplastic, thixotropic and compressible fluid. Strong temperature history dependence plays a prevailing role in the whole shutdown and restart process. Thus, waxy crude oils under flowing conditions correspond to the non-isothermal flow of a viscoplastic material with temperature-dependent rheological properties. Besides, the restart of a waxy crude oil is simulated by the isothermal transient flow of a weakly compressible thixotropic fluid in axisymmetric pipe geometry. We retain the Houska model to describe the thixotropic/viscoplastic feature of the fluid and compressibility is introduced in the continuity equation. The viscoplastic constitutive equation is involved using an augmented Lagrangian method and the resulting equivalent saddle-point problem is solved thanks to an Uzawa-like algorithm. Governing equations are discretized using a Finite Volume method and the convection terms are treated thanks to a TVD (Total Variation Diminishing) scheme. The Lagrangian functional technique usually used for incompressible viscoplastic flows, is adapted to compressible situations. Several numerical results attest the good convergence properties of the proposed transient algorithm. The non-isothermal results highlight the strong sensitivity of

  3. The pros and cons about restarting and awareness about nuclear power generation. Further findings from INSS's analysis of the opinion survey answers

    International Nuclear Information System (INIS)

    In this paper, the pros and cons and the awareness of their background on the restart of nuclear power plants (NPPs) were analyzed based on the data of opinion polls conducted by news organizations and INSS. The results were as follows: (1) Although opposition to restart has been nearly 60% in the case when the question has only 2 choices of pros and cons, the 60% includes many people who “cannot say either way”. (2) For approval of restarting, it is necessary to have the attitude to tolerate the use of nuclear power generation (NPG), and it is extremely important that people think “safety has been confirmed,” but it does not seem to be enough reason to hesitate in the restart. (3) From the open-ended question about the influences of long-term shutdown of NPPs on the respondents and Japanese society, 50% of the respondents said there was no influence. 20% said they felt safe or easy because no accident could occur. Only 20% described economic influence, such as a real increase in electricity cost, especially among men or the people who were in favor of restarting. (4) When the above-described influences, people's recognition of the utility of NPG (or thinking about the problems that occur when reducing NPG) and people's criteria for selecting an electric power generation method were organized in terms of the 3Es (Energy security, Economic efficiency, Environment), there was quite a difference from the idea of energy policy that emphasizes a good balance among these 3Es. People's recognition of the utility of NPG has been declining. This was considered to be the reason that neither the present influences nor the importance of the invisible influence at a macro-level had been recognized. For restarting to be supported by many people, there is a need for appropriate information about not only the safety, but also the importance from the viewpoints of the 3Es and the benefits of NPG in the 3Es. (author)

  4. Sensitive assessment of the virologic outcomes of stopping and restarting non-nucleoside reverse transcriptase inhibitor-based antiretroviral therapy.

    Directory of Open Access Journals (Sweden)

    Anna Maria Geretti

    Full Text Available BACKGROUND: Non-nucleoside reverse transcriptase inhibitor (NNRTI-resistant mutants have been shown to emerge after interruption of suppressive NNRTI-based antiretroviral therapy (ART using routine testing. The aim of this study was to quantify the risk of resistance by sensitive testing and correlate the detection of resistance with NNRTI concentrations after treatment interruption and virologic responses after treatment resumption. METHODS: Resistance-associated mutations (RAMs and NNRTI concentrations were studied in plasma from 132 patients who interrupted suppressive ART within SMART. RAMs were detected by Sanger sequencing, allele-specific PCR, and ultra-deep sequencing. NNRTI concentrations were measured by sensitive high-performance liquid chromatography. RESULTS: Four weeks after NNRTI interruption, 19/31 (61.3% and 34/39 (87.2% patients showed measurable nevirapine (>0.25 ng/ml or efavirenz (>5 ng/ml concentrations, respectively. Median eight weeks after interruption, 22/131 (16.8% patients showed ≥1 NNRTI-RAM, including eight patients with NNRTI-RAMs detected only by sensitive testing. The adjusted odds ratio (OR of NNRTI-RAM detection was 7.62 (95% confidence interval [CI] 1.52, 38.30; p = 0.01 with nevirapine or efavirenz concentrations above vs. below the median measured in the study population. Staggered interruption, whereby nucleos(tide reverse transcriptase inhibitors (NRTIs were continued for median nine days after NNRTI interruption, did not prevent NNRTI-RAMs, but increased detection of NRTI-RAMs (OR 4.25; 95% CI 1.02, 17.77; p = 0.03. After restarting NNRTI-based ART (n = 90, virologic suppression rates <400 copies/ml were 8/13 (61.5% with NNRTI-RAMs, 7/11 (63.6% with NRTI-RAMs only, and 51/59 (86.4% without RAMs. The ORs of re-suppression were 0.18 (95% CI 0.03, 0.89 and 0.17 (95% CI 0.03, 1.15 for patients with NNRTI-RAMs or NRTI-RAMs only respectively vs. those without RAMs (p = 0.04. CONCLUSIONS

  5. CERREX Software Application at Budapest Research Reactor (BRR). Appendix V

    International Nuclear Information System (INIS)

    The BRR is a Russian designed WWRS-M10 reactor. It is a tank type, light water cooled and moderated research reactor. Its main goal is radioisotope production and neutron physics research. The BRR went critical in 1959 and during 52 years of operation, two modernizations and one partial decommissioning were carried out. After the second modernization, the reactor restarted in 1993. The planned lifetime is 30 years, and final shutdown will be in 2023. After the final shutdown, there will be a 2 year transition period and then partial decommissioning. The final goal is to dismantle the reactor system, most of the auxiliary systems and subsystems and hand the reactor building over to the Hungarian Academy of Sciences as an unrestricted site. The organization operating BRR has, since 2004, participated in the IAEA regional TC project RER/3/009 on Support in Planning for the Decommissioning of Nuclear Power Plants and Research Reactors. The first Preliminary Decommissioning Plan (PDP) was developed in 2005 and followed IAEA recommendations. A revised PDP utilized IAEA guidance given. The IAEA Expert Mission to discuss the decommissioning planning of the BRR was organized in 2010. The PDP database and the study of earlier decommissioning is the basis of CERREX software 'Inventory' and 'ISDC' work-sheets, despite the fact that the PDP structure is different from the CERREX structure

  6. Reactor building

    International Nuclear Information System (INIS)

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

  7. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

  8. Nature of radio feature formed by re-started jet activity in 3C 84 and its relation with γ-ray emissions

    Science.gov (United States)

    Nagai, H.; Chida, H.; Kino, M.; Orienti, M.; D'Ammando, F.; Giovannini, G.; Hiura, K.

    2016-02-01

    Re-started jet activity occurred in the bright nearby radio source 3C 84 in about 2005. The re-started jet is forming a prominent component (namely C3) at the tip of jet. The component has showed an increase in radio flux density for more than 7 years while the radio spectrum remains optically thin. This suggests that the component is the head of a radio lobe including a hotspot where the particle acceleration occurs. Thus, 3C 84 is a unique laboratory to study the physical properties at the very early stage of radio source evolution. Another important aspect is that high energy and very high energy γ-ray emissions are detected from this source. The quest for the site of γ-ray emission is quite important to obtain a better understanding of γ-ray emission mechanisms in radio galaxies. In this paper, we review the observational results from very long baseline interferometry (VLBI) monitoring of 3C 84 reported in series of our previous papers. We argue the nature of re-started jet/radio lobe and its relation with high-energy emission.

  9. Experimental RA reactor operation with 80% enriched fuel - Program of experimental operation: a) Program of experimental operation with 80% enriched fuel at low power, b) contents of the experimental operation with 80% enriched fuel at higher power levels

    International Nuclear Information System (INIS)

    Highly enriched (80%) uranium oxide fuel was regularly used in the mixed reactor core with the 2% enriched fuel since 1976. The most important changes related to reactor operation, in comparison with the original design project were related to reactor core fuelling schemes. At the end of 1979 reactor was shutdown due to the corrosion coating noticed on some fuel elements and due to decrease quality of the heavy water. Subsequently the Sanitary inspector of Serbia has prohibited further reactor operation. Restart of the reactor will not be a simple continuation of operation. It is indispensable to perform complete experimental program including measurements of critical parameters at different power levels for the core with fresh 80% enriched fuel. The aim of this document is to obtain working permission and its contents are in agreement with the procedure demanded by the Safety Committee of the Institute. It includes results of optimization and safety analysis for the initial reactor core. Since the permission for restart is not obtained, a separate RA reactor safety report is prepared in addition to the program for experimental operation. This report includes: detailed program for reactor experimental operation with 80% enriched fuel in the core at low power levels, and contents of the experimental operation with 80% enriched fuel in the core at higher power levels

  10. Plasma reactor

    OpenAIRE

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

    2008-01-01

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

  11. Reactor physics

    International Nuclear Information System (INIS)

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

  12. Neutron sensors in the SP-100 reactor control system

    International Nuclear Information System (INIS)

    The reference reactor control approach for the mature generic flight system (GFS) utilizes highly reliable and diverse reactor outlet temperature measurements for control and protection. Although system dynamic analyses demonstrated that this approach is satisfactory for various modes of operation (including transients involving failure or degradation of equipment), the use of a neutron monitoring system (NMS) for initial startup and for an early period of power operation has been studied to improve the performance of the reactor control design. Control strategies were developed, simulation analyses were produced, and stability margins were examined. In this updated control approach, the signals from the NMS are used for the initial startup, for restarts, for power range control, and for protection from overpower transients as long as reliable data is available from the NMS. The results show satisfactory performance for the updated controls. If the lifetime of the NMS is shorter than that of the flight system, the reactor control will revert to the reference control approach employing reactor outlet temperature measurements only

  13. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

    International Nuclear Information System (INIS)

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW

  14. Compact Reactor

    International Nuclear Information System (INIS)

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

  15. Fast reactor development programme in France during 1991

    International Nuclear Information System (INIS)

    In 1991, the total electric power consumption in Franc, was 374.5 TWh, a 7.1% increase was attained compared to 1990. The exported power was 53.4 TWh. 72.7% of power generated was produced by nuclear power plants. Two 1300 MW NPPs started industrial operation in 1990. During the whole year 1991, both Phenix and Superphenix were kept shut down. despite all efforts, it was not possible to give explanation of the negative reactivity incidents which occurred in Phenix three times in 1989 and once in September 1990, leading to automatic trips. In addition, tests were performed to assess and improve the situation in the case of total loss of coolant in all secondary circuits. Finally defects were detected in some welds of the secondary circuits and the repair is being carried out. Superphenix was kept shutdown for technical and political reasons. Among the technical reasons, was the construction of new fuel transfer station replacing the former fuel storage drum and the repair of the roof above one turbo generator that has collapsed in December 1990 under an usual weight of wet snow. On the political side, the decree for restarting Superphenix was cancelled by the 'Conseil d'Etat' because of a vice of form. France has given the highest priority to fulfil the requirements of the safety Authority in order to be able to restart both reactors

  16. 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

    Scandinavia). The nuclear renaissance is effective worldwide, with 33 power plants today under construction in the world and a lot of projects in discussion or in preparation in various countries (England, Italy, South Africa, USA...). In Europe, some countries, who phase-out the development nuclear energy, are also coming back in nuclear perspectives as Sweden, Italy, England, Poland,.. All these facts begin to give more work to the MTR (material testing reactors) for testing new materials and new fuels to improve their capacities and their performances. For the ZPR (Zero Power Reactors) test with new fuels allowing additives to suppress Bore utilisation, or allowing to reduce uranium consumption, will be necessary in the near future. For the safety dedicated reactors, test for compliance to last safety requirements are necessary. In this field the refurbishment of the CABRI reactor for Reactivity Insertion Accident studies, is now almost finished for test that should begin in 2010. For the radio isotope production the world demand is increasing year after year, especially for {sup 99}Mo, used in about 70 millions of medicine procedures each year in the world. Today 95% of this world production is assumed by five reactors: HFR (Netherlands), OSIRIS (France), SAFARI (South Africa), BRII (Belgium), and NRU (Canada). The youngest is OSIRIS (41 years) and should be close in 2015. Due to ageing problems NRU and HFR were shut down in 2009 for necessary repair. These points have conduced to some radio isotopes crisis in 2009. This paper explains some projects in line for the future to avoid this type of problems (FRMII initiative, RJH utilisation and PALLAS project). For training activities, needs are huge with nuclear renaissance, especially for the new countries coming back in nuclear field. It will also give a lot of opportunities to low power reactors and to the universities reactors. This paper also provides information on the status of the new projects such as the JHR ongoing

  17. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    International Nuclear Information System (INIS)

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code

  18. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Lisovyy, Oleksandr, E-mail: o.lisovyy@dimnp.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); Cherubini, Marco, E-mail: m.cherubini@ing.unipi.it [NINE, Via Livornese 1291, Pisa 56122 (Italy); Lazzerini, Davide, E-mail: d.lazzerini@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy)

    2015-03-15

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code.

  19. Physics aspects of reload and approach-to-critical of the NRU reactor after vessel repair

    International Nuclear Information System (INIS)

    The National Research Universal (NRU) reactor at Chalk River shut down on 2009 May 14 and there was a subsequent outage of 15 months to repair leaks from the vessel. On 2010 August 17, NRU returned to full power operation and resumed isotope production. This paper describes the physics aspects of reload, and the approach-to-critical (ATC) tests conducted to restart the reactor safely. Five ATC's, each at a different number of reloaded assemblies, plus a final one before reactor startup, were completed to confirm the calculated physics predictions of the subcritical state and critical point. Activities for preparation of the ATC tests, the responsibilities of the physicists during execution of the ATC's, and plots of neutron signal data during the ATC's are presented. The final measured critical point of CR 14 @190 cm agreed well with the calculated physics prediction of CR 14 @185 cm, or within ∼0.5 mk. (author)

  20. Evaluations of 105-K Reactor building floor slab frequencies and basemat stresses

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, R.L.

    1990-12-01

    This report describes the evaluation of natural frequencies of the concrete floor slabs in the K Reactor. The evaluation of the floor frequencies at elevation 34 ft and below is required to demonstrate that the floor frequencies of the Reactor Building are sufficiently high. This will provide confidence in the interim position that the structural amplification of the vertical motion are approximately in the same order or less than the structural amplification of the horizontal motion. Consequently, this evaluation determines the acceptability of using horizontal floor response spectra or two thirds of the horizontal floor response, on a case by case basis, as an interim estimate of the vertical motions in the K Reactor for restart.

  1. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

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

  2. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  3. Nuclear reactors

    International Nuclear Information System (INIS)

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

  4. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

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

  5. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

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

  6. Nuclear reactor

    International Nuclear Information System (INIS)

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

  7. Breeder reactors

    International Nuclear Information System (INIS)

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

  8. Status of Phenix operation and of sodium fast reactors in the world

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Martin, L. [Phenix plant, 30 - Bagnols sur Ceze (France); Courtois, C. [CEA Marcoule 30 (France)

    2007-07-01

    The French fast breeder reactor (FBR) Phenix restarted in 2003 after 6 years of safety reevaluation procedures. The goal of the experiments performed at Phenix is, first, to demonstrate the technical feasibility of transmutation of minor actinides and long-life products in a fast reactor and secondly, to acquire knowledge on structure materials for future energy systems and on innovative nuclear fuel concepts. After several years of Generation IV discussions, many countries have announced or confirmed their priority for the fast sodium reactor as a reference design. These countries today include Japan, China, Korea, India and Russia (simultaneously with lead reactors). The United States have announced a project for a waste-burning reactor. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision of building a prototype scheduled for operation in 2020. These declarations are all sustained in a very practical manner by ongoing events in this field. Following the excellent results obtained by the BN-600 (600 MWe), Russia has re-launched the BN-800 project. China is currently in the process of building a 75 MWt research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU (250 MWe) for divergence in 2008. In India, a 1200 MWt power reactor is under construction, scheduled for divergence in 2010, the first of 3 planned sodium reactors.

  9. LHC restart 2016

    CERN Multimedia

    CERN Audiovisual production

    2016-01-01

    Footage of first injection of proton beam in the LHC machine, that took place on Friday March 25, 2016, at the injection energy of 450 GeV. Interview to Mike Lamont, head of LHC operations and of Fabiola Gianotti, CERN Director General. General footage of the Cern Control Centre on first beam 2016 day, the LHC machine operators and Engineers in Charge at work, relevant screens summarizing the machine parametres. Views of the LHC tunnel and the 4 main experimental caverns with views of the ATLAS, CMS, ALICE and LHCb detectors. 3 D animations of the CERN accelerator complex, from the Linac, to the PS, SPS and the LHC. VOICE OVER : Alex Brown

  10. Restart the heart.

    Science.gov (United States)

    Klingberg, Karsten; Srivastava, David

    2016-01-01

    Early bystander cardiopulmonary resuscitation and rapid defibrillation are the most important factors for favourable outcomes after out of hospital cardiac arrest (OHCA)-as the new American Heart Association/European Resuscitation Council (AHA/ERC) guidelines emphasise. The patient in our case was a healthy young man who had a witnessed cardiac arrest due to a chest collision with the goalkeeper during a football match. Basic life support was immediately provided by his teammates until an automated external defibrillator was brought to the scene. Blunt cardiac injury (BCI) may result in injured myocardium or arrhythmias. Ventricular fibrillation due to BCI in absence of structural cardiac disease is one of the main causes of OHCA in young healthy athletes with high mortality rates. We demonstrate important aspects of the recently released guidelines on cardiac arrest and the chain of survival by the leading societies. PMID:26968838

  11. 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.

  12. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  13. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Development of Fast Breeder activities is being done mainly at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam and the total Scientific and Technical staff working at the Centre for development of FBRs is about 1200. The development work relating to the fuel fabrication and design and development for some of the fuel handling equipment is being done at the Bhabha Atomic Research Centre, Trombay, Bombay. Complete recovery from the fuel handling incident of FBTR was achieved during the beginning of 1989. Damaged guide tube and bent subassemblies were replaced, the incident was analysed in detail and appropriate remedial measures, viz., modifications in the fuel handling machine control logic and plug rotation logic were implemented to prevent its recurrence. Safety clearances for the restart of the reactor were obtained from the Atomic Energy Regulatory Board in May 1989. As steam generators were not valved in the secondary sodium system, the reactor power during this phase of operation was limited to 500 KWt. The main objectives during this phase were to complete the balance low power physics experiments and to operate the reactor for a sufficiently long time to assess the performance of various systems, in particular the neutronic instrumentation, control rod drive and safety logic system which were not in active service for the two years. From May to July, 1989, the reactor was successfully operated up to a power level of 500 KWt with 50% operating time. Design of PFBR is progressing intensively. (author). 1 tab

  14. Neutron guide system at the Budapest Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rosta, L.; Cser, L.; Grosz, T.; Toeroek, G. [Research Inst. for Solid State Physics, Budapest (Hungary); Belgya, T.; Molnar, G.; Revai, Z. [Institute of Isotopes, 1525 Budapest, KFKI, Pf. 49 (Hungary); Kaszas, G. [MIRROTRON Co. 1025 Budapest, Ozgida ut 19/3 (Hungary)

    1997-06-01

    The 10 MW research reactor was restarted after a full-scale refurbishment in 1993. An important task of the upgrading was the construction of a new experimental hall and three neutron guides have been installed for the instruments located in this area. The in-pile plug contains 25 x 100 mm{sup 2} section float glass optical elements coated with {sup 58}Ni. The guides of the same cross section have the following destination: NV1 has an interruption for the monochromator of a three-axis spectrometer and the prompt gamma activation analysis station is installed in end position. The second guide serves only for the small-angle scattering device and the third one for a reflectometer being constructed. (orig.).

  15. Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

    International Nuclear Information System (INIS)

    This report contains the findings from the first phase of a study on safety during outage and restart of nuclear power plants. Operational Readiness Verification (ORV) - in Swedish called Driftklarhetsverifiering (DKV) - refers to the test and verification activities that are necessary to ensure that plant systems are able to provide their required functions when needed - more concretely that all plant systems are in their correct functional state when the plant is restarted after an outage period. The concrete background for this work is that nine ORV related incidents were reported in Sweden between July 1995 and October 1998. The work reported here comprised a literature survey of research relevant for ORV issues, and an assessment of the present situation at Swedish NPPs with respect to ORV. The literature survey was primarily aimed at research related to NPPs, but also looked at domains where similar problems have occurred, such as maintenance in commercial aviation. The survey looked specifically for organisational and MTO aspects relevant to the present situation in Swedish NPPs. One finding was that ORV should be seen as an integral part of maintenance, rather than as a separate activity. Another, that there is a characteristic distribution of error modes for maintenance and ORV, with many sequence errors and omissions, rather than a set of unique error modes. An international study further showed that there are important differences in how procedures are used, and in the balance between decentralisation and centralisation. Several studies also suggested that ORV could usefully be described as a barrier system in relation to the flow of work, for instance using the following five stages: (1) preventive actions during maintenance/outage, (2) post-test after completion of work, (3) pre-test before start-up, (4) the start-up sequence itself, and (5) preventive actions during power operation - possibly including automatic safety systems. In the field survey

  16. Operational readiness verification, phase 1: A study on safety during outage and restart of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hollnagel, E. [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, V. [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

    2001-06-01

    This report contains the findings from the first phase of a study on safety during outage and restart of nuclear power plants. Operational Readiness Verification (ORV) - in Swedish called Driftklarhetsverifiering (DKV) - refers to the test and verification activities that are necessary to ensure that plant systems are able to provide their required functions when needed - more concretely that all plant systems are in their correct functional state when the plant is restarted after an outage period. The concrete background for this work is that nine ORV related incidents were reported in Sweden between July 1995 and October 1998. The work reported here comprised a literature survey of research relevant for ORV issues, and an assessment of the present situation at Swedish NPPs with respect to ORV. The literature survey was primarily aimed at research related to NPPs, but also looked at domains where similar problems have occurred, such as maintenance in commercial aviation. The survey looked specifically for organisational and MTO aspects relevant to the present situation in Swedish NPPs. One finding was that ORV should be seen as an integral part of maintenance, rather than as a separate activity. Another, that there is a characteristic distribution of error modes for maintenance and ORV, with many sequence errors and omissions, rather than a set of unique error modes. An international study further showed that there are important differences in how procedures are used, and in the balance between decentralisation and centralisation. Several studies also suggested that ORV could usefully be described as a barrier system in relation to the flow of work, for instance using the following five stages: (1) preventive actions during maintenance/outage, (2) post-test after completion of work, (3) pre-test before start-up, (4) the start-up sequence itself, and (5) preventive actions during power operation - possibly including automatic safety systems. In the field survey

  17. Reactor utilization

    International Nuclear Information System (INIS)

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

  18. Reactor Neutrinos

    CERN Document Server

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

    2005-01-01

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

  19. Nuclear reactors

    International Nuclear Information System (INIS)

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

  20. Nuclear reactors

    International Nuclear Information System (INIS)

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

  1. Reduction of fuel enrichment of the WWR-K research reactor (Kazakhstan)

    International Nuclear Information System (INIS)

    The WWR-K research reactor (rated power up to 10 MW) was put in operation at the Institute of Nuclear Physics of Kazakh Academy of Sciences in November, 1967. Great number of the vertical experimental channels for material irradiation and radioisotope production disposed both in the reactor core and in the reactor tank, as well as five horizontal radial channels and the single tangent one, allowed to solve a lot of scientific or industrial tasks. The reactor operated till October 1988; when, by decision of the regulatory body of the former USSR, the reactor was shut down from the necessity of performing a number of improvements to its safety because of high seismicity of the reactor site (more than 9 on the MK scale). Beginning with 1989, the stuff of the Kazakhstan National Nuclear Center fulfilled the works on enhancement of the reactor safety at the conditions of high seismicity. Essential assistance in these activities has been provided by IAEA and International Scientific Technical Center. As the result of these efforts (after estimation of the efficiency of the seismic safety up-grading reactor systems fulfilled by experts from Kazakhstan, Russia and IAEA), the decision permitting re-start of the reactor operation was taken by the regulatory body of the Kazakhstan - the Kazakhstan Atomic Energy Agency. It was stipulated reduction of the reactor power to 6 MW and creation more compact core configuration. The reactor was put in operation again in March 1998, and now at the reactor some research and irradiation works are being fulfilled

  2. Evaluation of the resilience of a full-scale down-flow hanging sponge reactor to long-term outages at a sewage treatment plant in India.

    Science.gov (United States)

    Onodera, Takashi; Takayama, Daisuke; Ohashi, Akiyoshi; Yamaguchi, Takashi; Uemura, Shigeki; Harada, Hideki

    2016-10-01

    Resilience to process outages is an essential requirement for sustainable wastewater treatment systems in developing countries. In this study, we evaluated the ability of a full-scale down-flow hanging sponge (DHS) reactor to recover after a 10-day outage. The DHS tested in this study uses polyurethane sponge as packing material. This full-scale DHS reactor has been tested over a period of about 4 years in India with a flow rate of 500 m(3)/day. Water was not supplied to the DHS reactor that was subjected to the 10-day outage; however, the biomass did not dry out because the sponge was able to retain enough water. Soon after the reactor was restarted, a small quantity of biomass, amounting to only 0.1% of the total retained biomass, was eluted. The DHS effluent achieved satisfactory removal of suspended solids, chemical oxygen demand, and ammonium nitrogen within 90, 45, and 90 min, respectively. Conversely, fecal coliforms in the DHS effluent did not reach satisfactory levels within 540 min; instead, the normal levels of fecal coliforms were achieved within 3 days. Overall, the tests demonstrated that the DHS reactor was sufficiently robust to withstand long-term outages and achieved steady state soon after restart. This reinforces the suitability of this technology for developing countries. PMID:27450993

  3. The RA nuclear research reactor at VINCA Institute as an engineering and scientific challenge

    International Nuclear Information System (INIS)

    The RA nuclear research at the Vinca Institute of Nuclear Sciences is the largest nuclear research facility in Yugoslavia and belongs to that generation of research reactors which have had an important contribution to nuclear technology development. As these older reactors were generally not built to specific nuclear standards, new safety systems had to be installed at the RA reactor for a renewal of its operating licence in 1984 and it was shut down, after 25 years of operation. Although all the required and several additional systems were built for the restart of the RA reactor, a disruption of foreign delivery of new control equipment caused its conversion to a 'dormant' facility, and it is still out of operation. Therefore, the future status of the RA reactor presents an engineering and scientific challenge to the engineers and scientists from Yugoslavia and other countries that may be interested to participate. To attract their attention on the subject, principal features of the RA reactor and its present status are described in detail, based on a recent engineering economic and safety evaluation. A comparative review of the world research reactors is also presented.(author)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

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

  5. Fast reactor development programme in France during 1995

    International Nuclear Information System (INIS)

    In 1995, the total amount of electricity produced in France was 471 TWh, out of which 358.2 TWh (76 %) were produced by nuclear power plants, 36.9 TWh (7.8 %) by conventional thermal plants, and 75.5 TWh (16 %) by hydraulic plants. The net electrical power consumption was 368.7 TWh. At the end of 1995, 'Electricite de France' had 54 PWR units in operation. The availability factor for these units was maintained at 81%. 1995 was marked by a decrease of unexpected shutdowns (1.8% in 1995 instead of 2.2% in 1994), a new reduction in programmed shutdown periods, and a good safety level was maintained. In the field of Fast Reactors, the main events of 1995 were the following. At the end of December 1994, the PHENIX reactor was authorized to perform its 49th cycle at 350 MW th (143 MWe). This 49th cycle was completed without any significant problems on April 7, 1995. During the remainder of the year, the reactor had been shut down in order to carry out several tasks within the scope of the ten-year extension of the PHENIX reactor's lifetime. Concerning the CREYS-MALVILLE plant (SUPER-PHENIX) the first part of the year was devoted to repairing argon leak of one of the IHX. Authorization to restart the reactor was given on August 22. The end of the year was beset by a number of minor incidents. The reactor was restarted at the end of 1995 and reactor power was increased by successive steps (30% Pn (Nominal Power) up to February 6 1996; followed by 50 %...). The 'Decret d'Autorisation de Creation' stipulates that because of its prototype character, SUPER PHENIX will have to be operated under conditions explicitly giving priority to safety and knowledge acquisition, with an objective of research and demonstration. In this context, the so-called 'knowledge acquisition' programme designed to prove the capacity of a large FBR to produce electricity on an industrial scale, to test the consumption of plutonium and minor actinides in a large fast reactor, as well as to provide

  6. Reactor physics recommissioning of Pickering NGS Units 1 and 2

    International Nuclear Information System (INIS)

    Investigations following the rupture of Pickering Unit 1 pressure tube G16 in 1983, led to the shutdown of Units 1 and 2 for pressure tube replacement and numerous other upgrades. They were recommissioned in 1987 and 1988 respectively. This paper surveys the procedures used during the reactor physics recommissioning of these two reactors and presents the results of these measurements. Special note is made of the differences between this recommissioning work, and the initial commissioning of new CANDU reactors. From a physics point of view, the restarted units differed substantially from the original design. The main difference in the core configuration involved the conversion of 10 of the original adjuster rods into shutoff rods. The reactivities of the remaining adjusters were increased. These substantial changes to the core, together with the full core of fresh fuel, necessitated a complete set of reactor physics recommissioning experiments. Some of our procedures differed from those used to commission a new reactor. This was due mainly to the high levels of tritium in the moderator D2O and to radiological hazards on the reactivity deck. Also, the high residual activities in the rebuilt cores lead to increased difficulties in neutron monitoring and higher subcritical neutron count rates (hence a higher than usual reactor power at first criticality). In general the results of our recommissioning measurements closely matched the results of presimulations using the OHRFSP and SMOKIN computer codes. Results for Unit 2 were generally better than those for Unit 1. This was due to improved procedures which resulted from our experiences with Unit 1. (author). 4 tabs., 9 figs

  7. RA Research reactor, Annual report 1970 - Operation and maintenance

    International Nuclear Information System (INIS)

    During 1970, the RA Reactor was operated at nominal power of 6.5 MW for 160 days, and 40 days at lower power levels. Total production mounted to 25968 MWh which is 3.87% higher than planned. The action plan was changed compared to the previous years because of sending the heavy water to France for re-concentration. Isotopic concentration of the heavy water was decreased to 99.05% and now after re-concentration it is 99.96%. Discrepancy from the action plan, in September was caused by the delay return of the heavy water for administrative and transportation difficulties. The restart of the reactor in September was postponed because the cladding of one fuel element was damaged immediately after the start-up, and the reactor had to be shutdown. In October and November reactor was in operation 28 and 25 days respectively which enabled to make up for the lost time. Reactor was used for irradiation and experiments according to the demand of 390 users, 340 from the Institute and 50 external users. This report contains detailed data about reactor power and experiments performed in 1969. It is concluded that the reactor operated successfully according to the plan. Shorter interruptions were caused only by difficulties with water supply pipes and sliding of the soil. Reactor was only twice scram shutdown because of the false signals caused by failures of the electronic control instrumentation. the period when reactor was not in operation was used for inspection of the reactor vessel internals. By using special TV cameras and telescopes, it was found that the there are no signs of corrosion on the reactor vessel, e.e. that the internals are in a very good state. Simultaneously, connection for the pipes of future emergency core cooling system were constructed. During 1970, the spent fuel was repacked from fuel channels into special aluminium casks. Four casks containing 660 fuel slugs was deposited int the storage pool No.4. There is now 18 casks with 2951 spent fuel slugs in

  8. Reactor container

    International Nuclear Information System (INIS)

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

  9. Reactor building

    International Nuclear Information System (INIS)

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

  10. Nuclear reactors

    International Nuclear Information System (INIS)

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

  11. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

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

  12. Nuclear reactor

    International Nuclear Information System (INIS)

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

  13. SRS reactor stack plume marking tests

    International Nuclear Information System (INIS)

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart

  14. Nuclear reactor

    International Nuclear Information System (INIS)

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

  15. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    International Nuclear Information System (INIS)

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart

  16. Conditions of the fluid in the RHRS pipe line after the loss and restart of the RHRS pump

    International Nuclear Information System (INIS)

    Due to a recent event at pressurized water reactor (PWR), the impact of a loss of shutdown cooling during reduced RCS inventory conditions was evaluated. The evaluation for the loss of RHRS assumed loss of shutdown cooling initially and then the RHRS pumps were restored. This analysis is based upon a decay heat rate corresponding to 2.66 days following reactor shutdown with an initial liquid level at the elevation of the reactor flange. When the RHRS pump was restored, the calculation revealed that a void fraction of 10-50% could exist in the hot leg pipe where the RHRS pumps take suction. The study presented in this paper was conducted to determine the maximum voids permissible in the hot leg piping and still ensure that the water entering the RHRS pump did not have steam voids. RELAP5 was used to perform this evaluation. RELAP5/M3.3 is a computer code used for the thermal-hydraulic analysis of transients and small-break accidents and transients in light-water nuclear power plants. It uses a one-dimensional, two-fluid model, consisting of steam and water, with the possibility of the vapor phase containing a non-condensable component. A model of the piping was developed and the conditions presented above were analyzed and the fluid conditions were reviewed. The results of the RELAP5 analyses indicated that for voids 30% or less in the hot leg, the RHRS pump would have adequate suction flow to remain in a stable operational condition. (author)

  17. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  18. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

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

  19. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 2: Sequoyah Unit 2 Cycle 3

    International Nuclear Information System (INIS)

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The other two benchmark critical calculations were the beginning-of-cycle (BOC) startup at both hot, zero-power (HZP) and HFP critical conditions. These latter calculations were used to check for consistency in the calculated results for different burnups and downtimes. The keff results were in the range of 1.00014 to 1.00259 with a standard deviation of less than 0.001

  20. Terminal-shock and restart control of a Mach 2.5, axisymmetric, mixed compression inlet with 40 percent internal contraction. [wind tunnel tests

    Science.gov (United States)

    Baumbick, R. J.

    1974-01-01

    Results of experimental tests conducted on a supersonic, mixed-compression, axisymmetric inlet are presented. The inlet is designed for operation at Mach 2.5 with a turbofan engine (TF-30). The inlet was coupled to either a choked orifice plate or a long duct which had a variable-area choked exit plug. Closed-loop frequency responses of selected diffuser static pressures used in the terminal-shock control system are presented. Results are shown for Mach 2.5 conditions with the inlet coupled to either the choked orifice plate or the long duct. Inlet unstart-restart traces are also presented. High-response inlet bypass doors were used to generate an internal disturbance and also to achieve terminal-shock control.

  1. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

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

  3. 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.)

  4. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

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

  5. Nuclear reactor

    International Nuclear Information System (INIS)

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

  6. Hybrid adsorptive membrane reactor

    Science.gov (United States)

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

    2011-01-01

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

  7. Hybrid adsorptive membrane reactor

    Science.gov (United States)

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

    2011-03-01

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

  8. Reactor container

    International Nuclear Information System (INIS)

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

  9. Restart of the Armenia-2 Nuclear Power Station: Radiological emergency preparedness considerations for the nearby American community

    International Nuclear Information System (INIS)

    The Armenia Nuclear Power Station is located at Metsamor, approximately 30 km NW of the capital, Yerevan. The station, a two-unit, first-generation Soviet-designed VVER-440/270 pressurized water reactor plant was closed following the 1988 earthquake near Spitak. Because of a severe energy shortage the Government of Armenia has undertaken a program to recommission Unit 2. The plant design and circumstances surrounding its closure caused members of the U.S. Embassy staff and the American community in Armenia to express concerns for their safety in the event of a radiological emergency. In response, two representatives from the U.S. Department of Energy's International Nuclear Safety Program traveled to Armenia to review the Status of radiological emergency preparedness, meet with the American community, and make protective action recommendations. In this presentation we examine the major issues associated with recommissioning of Armenia-2, the challenges involved with developing a radiological emergency preparedness program for the American community, and our recommendations for protective actions in the absence of a strong communications and radiological monitoring infrastructure

  10. CANDU load following test in ICN research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbas, S.; Palleck, S. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Hohoianu, G.; Parvan, M. [Inst. for Nuclear Research, ICN (Romania)

    2008-07-01

    To study the performance of CANDU fuel under Load Following (LF) conditions, a CANDU 6 type fuel element with 8.0 wt% {sup 235}U enrichment was irradiated in the 14 MW TRIGA materials testing reactor at the Institute for Nuclear Research (ICN) in Romania. This experiment was developed under the INR-AECL Memorandum for Co-operation in research and development of nuclear energy and technology. The fuel element underwent a successful demonstration of LF capability, where the fuel element withstood 200 daily cycles from 27 to 54 kW/m (average element linear power), as well as additional ramps due to reactor trips and restarts during the test period. The fuel element underwent a series of post-irradiation destructive and nondestructive examinations after the LF test irradiation. No performance or integrity issues were observed. This paper presents a description of the test facility, details of the test irradiation conditions and the post-irradiation examination results with discussion on their relation to CANDU fuel performance and integrity. (author)

  11. 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)

  12. Department of reactor technology

    International Nuclear Information System (INIS)

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

  13. RB reactor noise analysis

    International Nuclear Information System (INIS)

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

  14. Analysis of core damage frequency due to external events at the DOE [Department of Energy] N-Reactor

    International Nuclear Information System (INIS)

    A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs

  15. Analysis of core damage frequency due to external events at the DOE (Department of Energy) N-Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lambright, J.A.; Bohn, M.P.; Daniel, S.L. (Sandia National Labs., Albuquerque, NM (USA)); Baxter, J.T. (Westinghouse Hanford Co., Richland, WA (USA)); Johnson, J.J.; Ravindra, M.K.; Hashimoto, P.O.; Mraz, M.J.; Tong, W.H.; Conoscente, J.P. (EQE, Inc., San Francisco, CA (USA)); Brosseau, D.A. (ERCE, Inc., Albuquerque, NM (USA))

    1990-11-01

    A complete external events probabilistic risk assessment has been performed for the N-Reactor power plant, making full use of all insights gained during the past ten years' developments in risk assessment methodologies. A detailed screening analysis was performed which showed that all external events had negligible contribution to core damage frequency except fires, seismic events, and external flooding. A limited scope analysis of the external flooding risk indicated that it is not a major risk contributor. Detailed analyses of the fire and seismic risks resulted in total (mean) core damage frequencies of 1.96E-5 and 4.60E-05 per reactor year, respectively. Detailed uncertainty analyses were performed for both fire and seismic risks. These results show that the core damage frequency profile for these events is comparable to that found for existing commercial power plants if proposed fixes are completed as part of the restart program. 108 refs., 85 figs., 80 tabs.

  16. Evaluation of a hybrid anaerobic biofilm reactor treating winery effluents and using grape stalks as biofilm carrier.

    Science.gov (United States)

    Wahab, Mohamed Ali; Habouzit, Frédéric; Bernet, Nicolas; Jedidi, Naceur; Escudié, Renaud

    2016-07-01

    Wine production processes generate large amount of both winery wastewater and solid wastes. Furthermore, working periods, volumes and pollution loads greatly vary over the year. Therefore, it is recommended to develop a low-cost treatment technology for the treatment of winery effluents taking into account the variation of the organic loading rate (OLR). Accordingly, we have investigated the sequential operation of an anaerobic biofilm reactor treating winery effluents and using grape stalks (GSs) as biofilm carrier with an OLR ranging from 0.65 to 27 gCOD/L/d. The result showed that, during the start-up with wastewater influent, the chemical oxygen demand (COD) removal rate ranged from 83% to 93% and was about 91% at the end of the start-up period that lasted for 40 days. After 3 months of inactivity period of the reactor (no influent feeding), we have succeeded in restarting-up the reactor in only 15 days with a COD removal of 82% and a low concentration of volatile fatty acids (1 g/L), which confirms the robustness of the reactor. As a consequence, GSs can be used as an efficient carrier support, allowing a fast reactor start-up, while the biofilm conserves its activity during a non-feeding period. The proposed hybrid reactor thus permits to treat both winery effluents and GSs. PMID:26652186

  17. Fusion reactors-high temperature electrolysis (HTE)

    International Nuclear Information System (INIS)

    Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 18000C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 14000C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 106 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

  18. Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

    International Nuclear Information System (INIS)

    significantly lower enthalpies than those required for melting, when the fuel burnup exceeds approximately 40 MWd/kgU. This issue is investigated by reviewing all high-burnup UO2 fuel rods that have failed in RIA simulation tests in the Japanese Nuclear Safety Research Reactor and the French CABRI pulse reactor to date. Data from thirteen failed rods, with burnups between 44 and 64 MWd/kgU, indicate that clad tube failure does not necessarily lead to fuel pellet dispersal. In fact, the data suggest that a peak fuel radial average enthalpy of at least 500 J/gUO2 is required to expel a significant part (>10 %) of the fuel inventory into the coolant. However, this empirical enthalpy threshold for fuel dispersal from high-burnup fuel rods cannot be directly applied to light water reactors, since the power pulses and/or the cooling conditions used in the pulse tests differ notably from those expected in LWRs under RIA

  19. Accelerated Proximal Gradient Method Based on Restart Technique%基于重启技术的加速邻近梯度算法

    Institute of Scientific and Technical Information of China (English)

    赵静

    2014-01-01

    The accelerated proximal gradient method is upgraded from gradient method. Although the efficiency of the accelerat-ed proximal gradient method is apparently better than that of the gradient method, the trajectory of the accelerated proximal gra-dient method may oscillate. To overcome this drawback, restart techniques are applied to the accelerated proximal gradient meth-od. Numerical examples show that the techniques are useful.%加速邻近梯度法是在梯度法基础上的一个改进,虽然效率比梯度法有明显提高,但仍存在收敛轨迹出现往回迭代的情况。为了克服该缺点,提出了把重启技术应用在加速邻近梯度法上的方法,并通过数值例子进行了比对,证明了该技术的有效性。

  20. Reactor Physics Training

    International Nuclear Information System (INIS)

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

  1. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

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

  2. Control methods for reduction of DC link capacitor and restarting at instantaneous power failure in PWM converter. Den prime atsugata PWM converter no heikatsu condenser yoryo teigen oyobi shuntei saishido seigyoho

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, K.; Endo, T.; Honbu, M.; Nando, K. (Hitachi, Ltd., Tokyo (Japan))

    1992-01-15

    With an objective to realize a practical use of the voltage type converter, discussions were given on a method to reduce capacity of a smoothing capacitor and a method to restart it at an instantaneous power failure. Their control characteristics were evaluated in a total digitalization experiment using a one -chip micro computer and a DSP. These methods are aimed at reducing sizes of converter equipment. The following knowledge was obtained from the discussions: In the method to reduce the smoothing capacitor capacity, considerations were given on a system to detect in a direct current amount and at a high speed the inverter input current from the motor current and the PWM signal pulse, which was used as the base to reduce the direct current voltage at a sudden load change by compensating the load disturbance. When the carrier frequencies in the converter and the inverter were made to agree with each other and the signal phase difference is changed from 0{degree} to 180{degree},the capacitor current was reduced by (1/3). In the method of restarting at an instantaneous power failure, the restarting was verified stable by setting the lower limit lower than the diode rectification voltage, and switching it to the diode rectification when the direct current voltage is lower than this value. 6 refs., 18 figs.

  3. 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

  4. 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.)

  5. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  6. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  7. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest. PMID:26680444

  8. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  9. Exploitation continuation of Fessenheim nuclear plant nr 1 reactor after thirty years of operation

    International Nuclear Information System (INIS)

    After having recalled the regulatory framework, this report indicates how the Fukushima accident has been taken into account by the French nuclear safety authority (ASN) for the decision of keeping on operating the Fessenheim nuclear plant. Then, after a general presentation of nuclear installations, the report describes some peculiarities of the Fessenheim power plant with respect to the other French nuclear plants. It comments and discusses various issues: reactor exploitation, fuel management, vessel exploitation, exploitation of the main secondary circuits, of the confinement enclosure, and of other equipment. It recalls significant events, exploitation rules, and modifications brought to the reactor. It gives a global assessment. The authors report the safety re-examination (approach, compliance examination, security re-assessment), controls performed during decennial inspection (main controls and tests, implementation of modifications foreseen by safety re-examination, significant events, monitoring by the ASN, reactor restarting after the third decennial inspection). Perspectives are then discussed for the ten following years in terms of maintenance policy, ageing management, reactor vessel serviceability, and additional actions within the frame of ageing management. The operation continuation is then discussed

  10. Nuclear reactor building

    International Nuclear Information System (INIS)

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

  11. Creys-Melville: A case study on the acceptability of fast breeder reactors

    International Nuclear Information System (INIS)

    The Operator of CREYS-MALVILLE, being convinced that in a democratic country like France, Nuclear Energy in general and fast reactors in particular may only be developed in agreement with public opinion, will continue to develop its external Communication - without passion but with conviction. This paper has recalled the interests at stake associated with the resumption of power at CREYS-MALVILLE, the media and political contexts which have followed the construction and operation of this Plant, the administrative resources, and the communication equipment which have been developed over these years. All these means would be wasted and all communication efforts useless if they did not rely on the perseverance and competence of the 650 technicians and engineers guarantors of the Safety at CREYS-MALVILLE, for whom the coming Public Enquiry will be yet another occasion to explain, to France and its neighbours, the necessity to restart SUPERPHENIX

  12. Reactor Physics Programme

    International Nuclear Information System (INIS)

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

  13. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10-9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  14. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  15. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    There are still many problems in the field of measurement and control of neutron flux. The present studies in connexion with high flux reactors contribute to the solution of these problems which concern specialists in reactor control. The present state of this investigation and the results of different studies carried out in France by the C A and the EDF are pointed out: A - In the nuclear instrumentation field, work is at present devoted to the technologies used to develop detectors and cables, which have to work at high temperature and in a high γ background; fast electronic techniques are applied to fission counters to measure low neutron fluxes in a high γ background (10 Rh). B - In the control and safety field, there is a real need for studies on the behaviour of reactors in the subcritical state. This increases the margin of security during restarts when poison effects must be overcome The perturbations due to control rod movements necessitate a new organisation of power level safety and control assemblies, in connexion with thermal or activation measurements. Two methods of fast start-up are described. They are related to the fission rate measurement as a function of time. This is done either continuously by a constant and high reactivity change, or step by step. The application of automatic techniques to detector motion seems to give the answer to control and safety in normal start-up. C - The scope of these studies covers the methods used for the control of E.D.F. 3, which are described. (authors)

  16. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

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

  17. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

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

  18. Process heat reactors

    International Nuclear Information System (INIS)

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

  19. Reactor System Design

    International Nuclear Information System (INIS)

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

  20. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

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

    2009-08-01

    data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of

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

    International Nuclear Information System (INIS)

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

  2. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

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

  3. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  4. High solids fermentation reactor

    Science.gov (United States)

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

    1993-01-01

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

  5. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

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

  6. Fusion reactor studies

    International Nuclear Information System (INIS)

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

  7. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  8. Support Required for Safety Management of Research Reactors, Especially Those in Extended Shutdown

    International Nuclear Information System (INIS)

    The Japan Materials Testing Reactor (JMTR) was operated for 38 years, with 165 cycles for various users, from its first criticality in 1968. The JMTR has been in an extended shutdown period since August 2006 for refurbishment and will restart in 2011. The JMTR is the only testing reactor dedicated to irradiation testing of materials and fuels in Japan. In response to the strong requests of the various users, the Japan Atomic Energy Agency (JAEA) decided to refurbish the JMTR for long term operation, and the work started in 2007. In order to improve the reliability and safety of the reactor, a number of reviews and inspections were conducted in parallel with the refurbishment. The experience of the JMTR during the period could be shared with other organizations which currently or potentially have a similar situation. A periodic safety review was carried out to confirm the integrity of the JMTR facilities, and a 10 year maintenance plan was developed in 2004. Before the restart of the JMTR, equipment to be renewed was selected based on an evaluation of its damage and wear in terms of ageing, significance to safety functions and past safety related maintenance, in order to enhance the operational capability. Renewal work on the power supply system, boiler, radioactive waste facility, etc., was finished as scheduled. Work on the reactor control system, the nuclear instrumentation system and the primary cooling system is being carried out. As for safety management during reactor operation, the owner's periodic and daily inspections are supposed to be carried out in order to maintain the integrity and reliability of the facility. The performance of the facilities and equipment is confirmed through inspections. During the extended shutdown period, a special classification of the facilities was made based on whether their functions are required continuously during the refurbishment period. Maintenance work and periodic inspections, including those by the regulatory

  9. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  10. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

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

  11. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  12. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  13. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-09-01

    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format.

  14. Shipment of 255 DIDO fuel elements to the Savannah River Site to empty the storage and reactor pools at Risoe National Laboratory

    International Nuclear Information System (INIS)

    The DR-3 reactor, owned and operated by the Danish National Laboratory, was built in the late 1950's and initiated operation in January 1960. At that time the DR-1 and DR-2 reactors were already in operation. The main purpose if of Danish research reactor DR-3 was material and fuel testing. Until 1989 the reactor utilized HEU fuel elements. Conversion to the LEU fuel cycle was accomplished in 1990. DOE restarted the return program of for Foreign Research Reactor fuel elements to the United States in 1994. From that time, through 1998, three IUO4 casks (one cask in 1994) operated by Transnucleaire (now named Cogema Logistics, ACL) were used to transport Risoe's fuel to the Savannah River Site (SRS) near Aiken, SC in the USA. In 1999, Risoe elected to issue a request for proposal to transport DR-3 the DIDO fuel elements to SRS with a new licensed cask designed to replace the IUO4 cask. ACL was awarded the contract to transport the irradiated fuel from DR-3 to SRS for the remainder of the FRR Fuel Return program (2009). However, on September 28, 2000, the Board of Governors of Risoe National Laboratory decided to shut down the Danish research reactor of DR3. There had been of R2 technical problems (corrosion on the aluminum reactor tank) and, due to anticipated increasing operational expenses, the Board elected to close the reactor facility. Shortly thereafter, the Danish Government asked the National Laboratory to empty the reactor and its reactor and storage pools containing a total of 255 Dido irradiated fuel elements and ship them to Savannah Rive Site. At that time, ACL was in the process of licensing the new TN-MTR package in the USA. The early shut down of the DR-3 reactor and consequently the resultant new shipping schedule was not compatible with ACL's equipment and licensing schedule for the cask. (author)

  15. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  16. A review of fast reactor program in Japan (April 2001 - March 2002)

    International Nuclear Information System (INIS)

    This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

  17. The Maple reactor project

    International Nuclear Information System (INIS)

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  18. High temperature reactors

    International Nuclear Information System (INIS)

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

  19. Spinning fluids reactor

    Science.gov (United States)

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

    2012-11-20

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

  20. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

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

  1. Regulatory analysis for the resolution of Generic Issue 94, ''Additional low-temperature overpressure protection for light-water reactors''

    International Nuclear Information System (INIS)

    Low-temperature overpressure protection (LTOP) is required in pressurized water reactors (PWRs) to provide protection against brittle reactor pressure vessel failure following an anticipated event. Typically these events are a result of either mass imbalance (excess charging in comparison to available letdown flow or inadvertent safety injection) or energy input transients (restarting an idle reactor coolant pump causing an increase in the reactor coolant system pressure as a result of mixing cold water from the inactive loop with the remainder of the hot fluid and as a result of direct energy addition from a warmer secondary side heat sink). The significance of these events is heightened during water-solid operations. Low-temperature overpressure protection is required in the shutdown modes of operation, Mode 4 -- Hot Shutdown, Mode 5 -- Cold Shutdown, and Mode 6 -- Refueling with the reactor vessel head bolted down. While operating in Modes 5 and 6 and with the reactor coolant temperature below 200 degree F, there are no technical specifications for containment integrity. The consequences of an unmitigated low-temperature overpressure event can be significant as a result of either containment bypass or failure of containment to isolate following reactor pressure vessel failure. This report presents the regulatory analysis for Generic Issue 94, ''Additional Low-Temperature Overpressure Protection for Light-Water Reactors.'' It includes (1) a summary of the issue, (2) the proposed technical resolution, (3) alternative resolutions considered by the Nuclear Regulatory Commission (NRC), (4) an assessment of the benefits and cost of the alternatives considered with additional emphasis on the recommended resolution, (5) the decision rationale, and (6) the impacts and relationships between GI-94 and other NRC programs and requirements. 35 refs., 60 tabs

  2. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  3. Visual inspections of N Reactor horizontal control rod channels

    International Nuclear Information System (INIS)

    Safety surveillance is performed in horizontal control rod (HCR) channels to locate conditions which could slow or block rod travel. The findings guide the application of preventive measures to assure eventual rod motion impairment will not occur. Borescopes and, more recently, miniaturized closed circuit television (CCTV) cameras have been used for these examinations. Inspections and measurement results are documented in annual surveillance reports, however reported CCTV observations have been limited to highlights. The objective of this report is to catalogue the CCTV recordings in a format suitable for analysis and interpretation and to ease the access to any desired location by noting tape counter readings corresponding with each tube block in view. Searching file tapes for conditions in a specific areas in the past required counting blocks as they passed the camera to determine the distance from a feature like the edge of the reflector or a steam vent gap. This report adds the observations from recent rod channel inspections (1987 and 1988) to a comprehensive survey of graphite conditions in the moderator and reflector regions of the N Reactor core. When completed, the stand-by status of graphite components will be available for use in restart or decommissioning deliberations

  4. Physics aspects of reload and approach-to-critical of the NRU reactor after vessel repair

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C.; Atfield, M.D.; Wang, X.; Nguyen, S.; Pfeiffer, P.; Budgell, J., E-mail: leungt@aecl.ca, E-mail: atfieldm@aecl.ca, E-mail: wangx@aecl.ca, E-mail: nguyens@aecl.ca, E-mail: pfeifferp@aecl.ca, E-mail: budgellj@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    The National Research Universal (NRU) reactor at Chalk River shut down on 2009 May 14 and there was a subsequent outage of 15 months to repair leaks from the vessel. On 2010 August 17, NRU returned to full power operation and resumed isotope production. This paper describes the physics aspects of reload, and the approach-to-critical (ATC) tests conducted to restart the reactor safely. Five ATC's, each at a different number of reloaded assemblies, plus a final one before reactor startup, were completed to confirm the calculated physics predictions of the subcritical state and critical point. Activities for preparation of the ATC tests, the responsibilities of the physicists during execution of the ATC's, and plots of neutron signal data during the ATC's are presented. The final measured critical point of CR 14 @190 cm agreed well with the calculated physics prediction of CR 14 @185 cm, or within ∼0.5 mk. (author)

  5. 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

  6. 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)

  7. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  8. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  9. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical conditions. The

  10. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  11. 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

  12. 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

  13. The nuclear soliton reactor

    International Nuclear Information System (INIS)

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

  14. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  15. 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

  16. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  17. 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

  18. PFBR reactor protection

    International Nuclear Information System (INIS)

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

  19. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  20. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  1. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  2. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  3. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  4. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  5. Reactor construction steels

    International Nuclear Information System (INIS)

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

  6. Commercialization of fast reactors

    International Nuclear Information System (INIS)

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

  7. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  8. Natural convection type reactor

    International Nuclear Information System (INIS)

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

  9. 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

  10. Reactor power control device

    International Nuclear Information System (INIS)

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

  11. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

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

  12. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  13. Remote robotic inspection of irregular surfaces on the inner diameter of the AECL NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zeller, B., E-mail: bzeller@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada); Lombardi, L., E-mail: llombardi@utex.com [Utex Scientific Instruments, Mississauga, Ontario (Canada); Cyr, P., E-mail: pcyr@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada); Mair, H.D., E-mail: dmair@utex.com [Utex Scientific Instruments, Mississauga, Ontario (Canada); Ginzel, R., E-mail: rginzel@eclipsescientific.com [Eclipse Scientific Ltd., Waterloo, Ontario (Canada)

    2013-01-15

    In May of 2009, the NRU (National Research Universal) reactor was forced to shut down after a small heavy water leak. In 2009-2010 repairs were performed in order to restart medical isotope production mid-August 2010. Since the NRU vessel's return to service, a series of periodic inspections is required to ensure the safe operation of the reactor. Eclipse Scientific in collaboration with Utex Scientific Instruments and Liburdi Automation developed the NDE inspection system for the In-Service Inspection program of the NRU vessel. In addition to the difficult environmental, delivery and inspection circumstances the inspection team was faced with the problem of doing an immersion inspection of the inside surface of the reactor vessel through a small 120 mm access port at a distance of more than 10 m to the inspection area at the bottom of the reactor. The vessel was built over 50 years ago and as the inner surface was modified by the repair program during the forced outage, there were no accurate drawings of the inner surface of the vessel that an automated system could rely upon. Eclipse Scientific in collaboration with Liburdi Automation developed a robotic arm designed to enter from the remote access port to deploy the Phased Array and Eddy Current Array inspection heads into the reactor vessel. The motion control and data acquisition system was developed in collaboration with Utex Scientific Instruments using their Inspection Ware software. This paper will highlight the challenges faced in the development of an inspection system capable of using ultrasonic signals to learn a surface and, using this acquired surface topography, effectively and safely deploy and articulate the different inspection heads required to perform the In-Service Inspection of the NRU vessel. (author)

  14. Remote robotic inspection of irregular surfaces on the inner diameter of the AECL NRU reactor

    International Nuclear Information System (INIS)

    In May of 2009, the NRU (National Research Universal) reactor was forced to shut down after a small heavy water leak. In 2009-2010 repairs were performed in order to restart medical isotope production mid-August 2010. Since the NRU vessel's return to service, a series of periodic inspections is required to ensure the safe operation of the reactor. Eclipse Scientific in collaboration with Utex Scientific Instruments and Liburdi Automation developed the NDE inspection system for the In-Service Inspection program of the NRU vessel. In addition to the difficult environmental, delivery and inspection circumstances the inspection team was faced with the problem of doing an immersion inspection of the inside surface of the reactor vessel through a small 120 mm access port at a distance of more than 10 m to the inspection area at the bottom of the reactor. The vessel was built over 50 years ago and as the inner surface was modified by the repair program during the forced outage, there were no accurate drawings of the inner surface of the vessel that an automated system could rely upon. Eclipse Scientific in collaboration with Liburdi Automation developed a robotic arm designed to enter from the remote access port to deploy the Phased Array and Eddy Current Array inspection heads into the reactor vessel. The motion control and data acquisition system was developed in collaboration with Utex Scientific Instruments using their Inspection Ware software. This paper will highlight the challenges faced in the development of an inspection system capable of using ultrasonic signals to learn a surface and, using this acquired surface topography, effectively and safely deploy and articulate the different inspection heads required to perform the In-Service Inspection of the NRU vessel. (author)

  15. Visual numerical steering in 3D AGENT code system for advanced nuclear reactor modeling and design

    International Nuclear Information System (INIS)

    Highlights: ► Numerical steering framework developed for deterministic neutron transport code AGENT to speed up the solution. ► Resulting speed up is on the order of 50%. ► Use of the steering framework is demonstrated modeling a TRIGA reactor. ► Numerical steering framework showed to be well suited for the deterministic neutron transport methods. - Abstract: The AGENT simulation system is used for detailed three-dimensional modeling of neutron transport and corresponding properties of nuclear reactors of any design. Numerical solution to the neutron transport equation in the AGENT system is based on the Method of Characteristics (MOCs) and the theory of R-functions. The latter of which is used for accurately describing current and future heterogeneous lattices of reactor core configurations. The AGENT code has been extensively verified to assure a high degree of accuracy for predicting neutron three-dimensional point-wise flux spatial distributions, power peaking factors, reaction rates, and eigenvalues. In this paper, a new AGENT code feature, a computational steering, is presented. This new feature provides a novel way for using deterministic codes for fast evaluation of reactor core parameters, at no loss to accuracy. The computational steering framework as developed at the Technische Universität München is smoothly integrated into the AGENT solver. This framework allows for an arbitrary interruption of AGENT simulation, allowing the solver to restart with updated parameters. One possible use of this is to accelerate the convergence of the final values resulting in significantly reduced simulation times. Using this computational steering in the AGENT system, coarse MOC resolution parameters can initially be selected and later update them – while the simulation is actively running – into fine resolution parameters. The utility of the steering framework is demonstrated using the geometry of a research reactor at the University of Utah: this new

  16. 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.)

  17. One piece reactor removal

    International Nuclear Information System (INIS)

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

  18. 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)

  19. The fusion reactor

    International Nuclear Information System (INIS)

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

  20. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

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

  1. 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)

  2. Gas-cooled reactors

    International Nuclear Information System (INIS)

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

  3. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

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

  4. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  5. Light water type reactor

    International Nuclear Information System (INIS)

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

  6. Naval propulsion reactors

    International Nuclear Information System (INIS)

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

  7. Iris reactor conceptual design

    International Nuclear Information System (INIS)

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

  8. 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

  9. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-06

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

  10. Physics of nuclear reactors

    International Nuclear Information System (INIS)

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

  11. Mirror reactor surface study

    International Nuclear Information System (INIS)

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

  12. 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.)

  13. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  14. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  15. Compact torsatron reactors

    International Nuclear Information System (INIS)

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

  16. Multi-purpose reactor

    International Nuclear Information System (INIS)

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

  17. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

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

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

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

  19. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

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

  20. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

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

  1. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

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

  2. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

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

  3. 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

  4. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

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

  5. Reactor engineering department annual report

    International Nuclear Information System (INIS)

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

  6. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  7. Reactor water sampling device

    International Nuclear Information System (INIS)

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

  8. Test reactor technology

    International Nuclear Information System (INIS)

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

  9. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

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

  10. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

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

  11. Reactor Sharing Program

    International Nuclear Information System (INIS)

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

  12. 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)

  13. Reactor de plasma

    OpenAIRE

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

    2008-01-01

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

  14. Integral nuclear reactor

    International Nuclear Information System (INIS)

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

  15. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

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

  16. Licensed operating reactors

    International Nuclear Information System (INIS)

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

  17. 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

  18. Course on reactor physics

    International Nuclear Information System (INIS)

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

  19. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

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

  20. Nuclear reactor theory

    International Nuclear Information System (INIS)

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

  1. PWR type reactor

    International Nuclear Information System (INIS)

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

  2. Fusion Reactor Materials

    International Nuclear Information System (INIS)

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

  3. International tokamak reactor

    International Nuclear Information System (INIS)

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

  4. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

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

  5. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-07-01

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

  6. nuclear reactor design calculations

    International Nuclear Information System (INIS)

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

  7. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

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

  8. Reactor monitoring system

    International Nuclear Information System (INIS)

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

  9. 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

  10. Risk and safety analysis in support of the operation at the High Flux Isotope Reactor at Oak Ridge

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a Level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10-4. It was dominated by flow blockages and loss of all AC power. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50 percent of the internal event initiated contribution and is dominated by seismic events. Several design and safety analysis studies were undertaken to support the restart of the HFIR. The first study involved a fracture mechanics analysis and redesign of the reactor operating conditions and safety system setting to provide a basis for future operation. Another study involved performing a risk analysis by combining the Level 1 PRA results with offsite consequence analyses under conservative assumptions about the fission product removal within the plant. Additional studies were performed to establish a long-term decay heat removal design basis. Finally, updated and upgraded loss-of-cooling accident studies were performed and are still underway. 5 refs., 11 figs., 6 tabs

  11. Risk and safety analysis in support of the operation at the High Flux Isotope Reactor at oak ridge

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of the discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a Level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10-4. It was dominated by flow blockages and loss of all AC power. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic events, wind, and fires. The overall external event contribution to core damage frequency is about 50 percent of the internal event initiated contribution and is dominated by seismic events. Several design and safety analysis studies were undertaken to support the restart of the HFIR. The first study involved a fracture mechanics analysis and redesign of the reactor operating conditions and safety system settings to provide a basis for future operation. Another study involved performing a risk analysis by combining the Level 1 PRA results with offsite consequences analysis under conservative assumptions about the fission product removal within the plant. Additional studies were performed to establish a long-term decay heat removal design basis. Finally, updated and upgraded loss-of-cooling accident studies were performed and are still underway

  12. Nuclear Power Reactor Wastes. A Regulatory Overview of the Current Situation in Argentina

    International Nuclear Information System (INIS)

    Since the middle of the 1980's up to the year 2005 the Argentinean nuclear power program was stagnated. Entering in 21st century the situation of the nuclear program had an inflection point as a consequence of the rise of the oil prices, the deep reduction of indigenous resources of natural gas and the increase of electricity demand that resulted in the growing of economical activity since 2001. At present the construction of Atucha II, the 3rd nuclear power reactor, has been restarted and negotiations for a 4th nuclear power reactor are in course. The only authorized site for final disposal of low level solid radioactive waste (AGE) suspended its operation at the end of the '90s and a new schedule for decision and commissioning of radioactive waste facilities were established by National Commission of Atomic Energy (CNEA). Nowadays, in the country, the radioactive waste (RW) from NPPs are stored in the same site where they are generated and it is not planned to have in operation a disposal facility before the second half of the next decade. So, the main technical regulatory concern during this period will be to assure that the storage of the radioactive waste is safe, the characterization of RW performed by Operators is appropriate and a reasonable recording system is implemented. (authors)

  13. Nuclear power reactor wastes. A regulatory overview of the current situation in Argentina

    International Nuclear Information System (INIS)

    Since the middle of the 1980's up to the year 2005 the Argentinean nuclear power program was stagnated. Entering in 21st century the situation of the nuclear program had an inflection point as a consequence of the rise of the oil prices, the deep reduction of indigenous resources of natural gas and the increase of electricity demand that resulted in the growing of economical activity since 2001. At present the construction of Atucha II, the 3rd nuclear power reactor, has been restarted and negotiations for a 4th nuclear power reactor are in course. The only authorized site for final disposal of low level solid radioactive waste (AGE) suspended its operation at the end of the '90s and a new schedule for decision and commissioning of radioactive waste facilities were established by National Commission of Atomic Energy (CNEA). Nowadays, in the country, the radioactive waste (RW) from NPPs are stored in the same site where they are generated and it is not planned to have in operation a disposal facility before the second half of the next decade. So, the main technical regulatory concern during this period will be to assure that the storage of the radioactive waste is safe, the characterization of RW performed by Operators is appropriate and a reasonable recording system is implemented. (author)

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

    International Nuclear Information System (INIS)

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

  15. A 2400 MWth liquid lead-cooled flexible conversion ratio (FCR) reactor

    International Nuclear Information System (INIS)

    The design of a 2400 MWth lead-cooled reactor with cores capable of accommodating both the need to burn legacy waste (conversion ratio of zero) as well as of sustaining a closed cycle (unity conversion ratio) is presented. The core employs the transuranic metallic fuel. The large pool vessel contains four intermediate heat exchangers (IHX) that couple the primary system to an efficient and compact supercritical CO2 power conversion system. To prevent CO2 from entering the core in case of intermediate heat exchanger tube rupture, a dual-free level design for the primary vessel is adopted. Ultimate decay heat removal is accomplished by passive means through an enhanced reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary cooling system (PSACS). The transient simulation of station blackout (SBO) using RELAP5-3D shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and PSACS by removing a small fraction of decay power with the PSACS. If a large fraction of total decay heat is removed by the PSACS, it becomes difficult to design the water-cooled PSACS to balance short-term and long-term decay heat removal requirements in unprotected accidents to satisfy peak cladding temperature limit and at the same time prevent reactor restart from overcooling. In case of scram, the proposed RVACS/PSACS approach can be designed to remove decay power even if the PSACS transfers a large fraction of the total decay power. Nevertheless, in a very unlikely scenario of SBO without scram there is ample time of ∼25 hours for an operator to manually scram the reactor. An SBO accident sequence framework was proposed for this range of responses. (authors)

  16. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  17. Reactor power control device

    International Nuclear Information System (INIS)

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

  18. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

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

  19. Reactor parameter simulation system

    International Nuclear Information System (INIS)

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

  20. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  1. New reactor type proposed

    CERN Multimedia

    2003-01-01

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

  2. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

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

  3. 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)

  4. Experience with Kamini reactor

    International Nuclear Information System (INIS)

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

  5. 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.)

  6. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

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

  7. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

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

  8. Reactor hot spot analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R.B.

    1985-08-01

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

  9. 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

  10. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

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

  11. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

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

  12. Pulsed fusion reactors

    International Nuclear Information System (INIS)

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

  13. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

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

  14. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

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

  15. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

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

  16. 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

  17. OECD Halden reactor project

    International Nuclear Information System (INIS)

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

  18. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

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

  19. Small reactor return

    International Nuclear Information System (INIS)

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

  20. Reactor lattice transport calculations

    International Nuclear Information System (INIS)

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

  1. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

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

  2. Future reactor experiments

    International Nuclear Information System (INIS)

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

  3. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

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

  4. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

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

  5. Generation IV reactors: economics

    International Nuclear Information System (INIS)

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

  6. Future reactor experiments

    Science.gov (United States)

    Wen, Liangjian

    2015-07-01

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

  7. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

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

  8. AVR reactor physics

    International Nuclear Information System (INIS)

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

  9. Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

    International Nuclear Information System (INIS)

    Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

  10. Moon base reactor system

    Science.gov (United States)

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

    1989-01-01

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

  11. BWR type nuclear reactor

    International Nuclear Information System (INIS)

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

  12. Emergency reactor scram system

    International Nuclear Information System (INIS)

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

  13. A modular reactor plant

    International Nuclear Information System (INIS)

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

  14. New fission reactor designs

    International Nuclear Information System (INIS)

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

  15. OECD Halden reactor project

    International Nuclear Information System (INIS)

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

  16. Reactor power measuring device

    International Nuclear Information System (INIS)

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

  17. Reactor safety engineering

    International Nuclear Information System (INIS)

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

  18. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

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

  19. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-28

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

  20. Materials for nuclear reactors

    International Nuclear Information System (INIS)

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