WorldWideScience

Sample records for edwin i. hatch-1 reactor

  1. Edwin I. Hatch Nuclear Plant, Unit 2. License application, general information

    International Nuclear Information System (INIS)

    1975-01-01

    An application for a direct cycle BWR is presented. The reactor will be located about 11 miles north of Baxley, Ga., near the south bank of the Ultamaha River in Appling County close to Hatch-1 Reactor. The core thermal power level will be 2537 MW(t) and the electrical power level will be 795 MW(e). Mechanical cooling towers will be utilized

  2. Aerial radiological survey of the area surrounding the Edwin I. Hatch Nuclear Plant, Baxley, Georgia

    International Nuclear Information System (INIS)

    Hilton, L.K.

    1978-11-01

    An airborne radiological survey of a 2146 km 2 area surrounding the Edwin I. Hatch Nuclear Plant was made 28-31 March 1977. Detected radioisotopes, and their associated gamma ray exposure rates, were consistent with that expected from the normal background emitters. Count rates observed at 152 m altitude are converted to equivalent exposure rates at 1 m above the ground, and are presented in the form of an isopleth map. Exposure rates measured with small portable instruments and soil sample analysis showed agreement with the airborne data

  3. Draft environmental statement: Related to operation of the Edwin I. Hatch Nuclear Plant Unit No. 2, Georgia Power Company: Docket No. 50-366

    International Nuclear Information System (INIS)

    1977-04-01

    The proposed action is the issuance of an operation license to the Georgia Power Company for the startup and operation of the Edwin I. Hatch Nuclear Plant, Unit No. 2 (Docket No. 50-366), located on the Altamaha River in Appling County, approximately 11 miles north from Baxley, Georgia. The information in this Statement represents the second, assessment of the environmental impact associated with the Edwin I. Hatch Nuclear Plant, Unit No. 2, pursuant to the guidelines of the National Environmental Policy Act of 1969 (NEPA) and 10 CFR Part 51 of the Commission's Regulations. After receipt of an application, in 1970, to construct this plant, the staff carried out a review of impact that would occur during the construction and operation of this plant. That evaluation was issued as a Final Environmental Statement in October 1972. As the result of that environmental review, a safety review, an evaluation by the Advisory Committee on Reactor Safeguards, and a public hearing in Baxley, Georgia and Washington, D.C., the AEC (now NRC) issued a permit in December 1972, for the construction of Unit No. 2 of the Edwin I. Hatch Nuclear Plant. As of February 1977, the construction of Unit No. 2 was 70% complete. With a proposed fuel-loading date of April 1978 for Unit No. 2, the applicant has petitioned for license to operate Unit No. 2 and has submitted (July 1975) the required safety and environmental reports to substantiate this petition. 97 refs., 18 figs., 37 tabs

  4. Technical evaluation of RETS-required reports for the Edwin I. Hatch Nuclear Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Young, T.E.; Magleby, E.H.

    1985-01-01

    A review of the reports required by federal regulations and the plant-specific Radiological Effluent Technical Specifications (RETS) for operations conducted during 1983 was performed. The periodic reports reviewed for the Edwin I. Hatch Nuclear Plant were the Annual Radiological Environmental Operating Report for 1983 and the Semiannual Radioactive Effluent Release Reports for 1983. The principal review guidelines were the plant's specific RETS, NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants'', and NRC Guidance on the Review of the Process Control Programs. The Licensee's submitted reports were found to be reasonably complete and consistent with the review guidelines. 7 refs

  5. Edwin I. Hatch Nuclear Plant, Unit 1. Semiannual operating report, September 12--December 31, 1974

    International Nuclear Information System (INIS)

    1974-01-01

    Hatch-1 achieved initial criticality on September 12, 1974, and initial synchronization on November 11, 1974. The unit is a BWR of 813 MW(e) and has generated 50,775.6 MWH since September 12 with the generator on line 865.62 hours. Net plant efficiency was 20.88 percent with plant availability of 28.32 percent. Information is presented concerning operations, changes, tests, safety related maintenance, primary coolant chemistry, occupational personnel radiation doses, and radioactive effluent releases. (U.S.)

  6. Edwin I. Hatch Nuclear Plant, Unit 1. Annual operating report, 1975

    International Nuclear Information System (INIS)

    1976-01-01

    Net electrical power generated was 3,102,479 MWh(e) with the reactor on line 6,159 hrs. Information is presented concerning power generation, shutdowns, corrective maintenance, reactor coolant chemistry tests, occupational radiation exposure, release of radioactive materials, reportable occurrences, and primary containment local leak rate tests

  7. Reactor oscillator - I - III, Part I

    International Nuclear Information System (INIS)

    Lolic, B.

    1961-12-01

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

  8. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  9. Mars expert Edwin, 17, amazes Euro.

    CERN Multimedia

    2001-01-01

    Edwin Kite represented the UK in the 'Life in the Universe' competition held at CERN, Geneva. In his presentation Could Mars Have Supported Advanced Life?, he presented models of the Martian atmosphere over thousands of millions of years and demonstrated how the Red Planet could have sustained algae-like life between 3 and 4 thousand million years ago (1/2 page).

  10. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  11. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  12. Edwin M. McMillan, A biographical sketch

    International Nuclear Information System (INIS)

    Lofgren, E.J.

    1994-07-01

    Edwin M. McMillan was one of the great scientists of the middle years of this century. He made notable contributions to nuclear, and particle physics, the chemistry of transuranic elements, and accelerator physics

  13. Reactor antineutrino detector iDREAM.

    Science.gov (United States)

    Gromov, M. B.; Lukyanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

    2017-09-01

    Industrial Detector for Reactor Antineutrino Monitoring (iDREAM) is a compact (≈ 3.5m 2) industrial electron antineutrino spectrometer. It is dedicated for remote monitoring of PWR reactor operational modes by neutrino method in real-time. Measurements of antineutrino flux from PWR allow to estimate a fuel mixture in active zone and to check the status of the reactor campaign for non-proliferation purposes. LAB-based gadolinium doped scintillator is exploited as a target. Multizone architecture of the detector with gamma-catcher surrounding fiducial volume and plastic muon veto above and below ensure high efficiency of IBD detection and background suppression. DAQ is based on Flash ADC with PSD discrimination algorithms while digital trigger is programmable and flexible due to FPGA. The prototype detector was started up in 2014. Preliminary works on registration Cerenkov radiation produced by cosmic muons were established with distilled water inside the detector in order to test electronic and slow control systems. Also in parallel a long-term measurements with different scintillator samples were conducted.

  14. Prometheus Reactor I and C Software Development Methodology, for Action

    International Nuclear Information System (INIS)

    T. Hamilton

    2005-01-01

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I and C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I and C Software Development Process Manual and Reactor Module Software Development Plan to NR for information

  15. Prometheus Reactor I&C Software Development Methodology, for Action

    Energy Technology Data Exchange (ETDEWEB)

    T. Hamilton

    2005-07-30

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

  16. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  17. Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2016-11-01

    The Edwin B. Newman Graduate Research Award is sponsored jointly by Psi Chi, the national honor society in psychology, and the APA. The award is presented annually to the psychology graduate student who submits the best research paper that was published or presented at a national, regional, or state psychological association conference during the past calendar year. The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and APA. Members of the 2016 Edwin B. Newman Award Committee were Shawn Carlton, PhD, Psi Chi representative; Christina Frederick-Recascino, PhD; John Norcross, PhD, APA representative; Karenna Malavanti, PhD, Psi Chi representative; Steven Kohn, PhD, Psi Chi representative; Warren Fass, PhD, Psi Chi representative; Chris Lovelace, PhD, Psi Chi representative; and Cathy Epkins, PhD, APA representative. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

  18. RA Reactor operation and maintenance (I-IX), Part I

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    The report on RA reactor operation and maintenance for year 1963 is divided in six tasks. This volume contains the introductory report, and three tasks of the final report, namely reactor exploitation, reactivity changes of the RA reactor before repair, planning of refuelling

  19. Neutronics in ICF reactor ''SENRI-I''

    International Nuclear Information System (INIS)

    Nakai, S.; Ido, S.; Yamanaka, C.

    1983-01-01

    The neutronic behavior of SENRI-I has been examined taking into account the effect of fuel rhoR and Pb tamper on the emitted neutron from micro-explosion. One dimensional neutron transport was calculated by ANISIN-JR code with the nuclear data GICX-40. The effect of beam ports on neutronics and neutron streaming was examined by the three dimensional Monte-Carlo calculation code MORSE-E with the same nuclear data. The emitted neutrons are softened noticeably by the increase of the compressed fuel rhoR and the thickness of Pb coating. The latter also multiplies the net neutron number from pellet. The energy deposition and temperature increase and its distribution in the blankets and structural elements were obtained as a function of neutron spectrum from pellet. As for the tritium breeding ratio, the softening of neutron has little effect because the decrease of breeding by 7 Li with softening is compensated by the increase of breeding by 6 Li. The breeding ratio was 1.678, 1.639 and 1.576 with 14 MeV neutron, rhoR=0.7, rhoR=3 and rhoR=6 respectively. Neutron shielding and streaming from beam ports were examined and the dose rate of final optical elements were calculated to estimate the life of mirror. All these results show the feasibility of SENRI-I as a long life, maintenance free ICF pulse reactor and motivate to go further investigation and design studies in detail. (author)

  20. Finding the Right Formula: Edwin H. Walker Jr

    Science.gov (United States)

    Keels, Crystal L.

    2005-01-01

    Edwin H. Walker Jr earned his doctorate in chemistry at age 27 and has barely looked back. With 13 publications under his belt before coming out of graduate school, he has also given more than 20 poster presentations in national venues, most recently at the American Chemical Society. He can also include securing a half-million-dollar National…

  1. Edwin L. Herr: Preeminent Scholar, Leader, Advocate, and Mentor

    Science.gov (United States)

    Engels, Dennis W.

    2012-01-01

    This profile celebrates and chronicles selected themes and highlights of the ideas, scholarly accomplishments, leadership, humanity, and work ethic of Edwin L. Herr, one of the major forces in the counseling profession, for purposes of archiving elements of his history and stimulating continuity of his ideas, achievements, and dedication.

  2. Pathos in Criticism: Edwin Black's Communism-as-Cancer Metaphor

    Science.gov (United States)

    Condit, Celeste M.

    2013-01-01

    Edwin Black's essay on "The Second Persona," introduced to rhetorical critics a rationale and model for a type of ideological criticism. Because it ignored the role of pathos in both the rhetoric Black purported to critique and in the construction of his own audience, Black's essay mis-described key features of Robert Welch's "Blue Book", which…

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-12-15

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

  4. Advanced Carbothermal Electric Reactor, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  5. Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2017-12-01

    The Edwin B. Newman Graduate Research Award is sponsored jointly by Psi Chi, the national honor society in psychology, and the APA. The award is presented annually to the psychology graduate student who submits the best research paper that was published or presented at a national, regional, or state psychological association conference during the past calendar year. The Edwin B. Newman Graduate Research Award was established in 1979. The award was established to recognize young researchers at the beginning of their professional lives and to commemorate both the 50th anniversary of Psi Chi and the 100th anniversary of psychology as a science (dating from the founding of Wundt's laboratory). It was named for Dr. Edwin B. Newman, the first national president of Psi Chi (1929) and one of its founders. He was a prolific researcher and a long-time chair of the Department of Psychology at Harvard University. Newman was a member of APA's Board of Directors, served as recording secretary of the board from 1962 to 1967, and was parliamentarian for the APA Council of Representatives for many years. He served both Psi Chi and APA in a distinguished manner for half a century. The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and APA. Members of the 2017 Edwin B. Newman Award Committee were Shawn Carlton, PhD, Psi Chi representative; Christina Frederick-Recascino, PhD; John Norcross, PhD, APA representative; Karenna Malavanti, PhD, Psi Chi representative; Steven Kohn, PhD, Psi Chi representative; Warren Fass, PhD, Psi Chi representative; Chris Lovelace, PhD, Psi Chi representative; and Cathy Epkins, PhD, APA representative. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  6. Twenty years of Triga Mark I reactor use

    International Nuclear Information System (INIS)

    Stasiulevicius, R.; Maretti Junior, F.

    1981-01-01

    This work is a report on the 20 years of activities of the Triga Mark I, research reactor located in Belo Horizonte, Brazil. It contains also a list of publications, details of operation and improvements introduced in the reactor as well as some perspectives for its future. (A.C.A.S.)

  7. Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

    International Nuclear Information System (INIS)

    Harper, J.R.; Garde, R.

    1981-11-01

    The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

  8. Applicability of the modified Emergency Department Work Index (mEDWIN at a Dutch emergency department.

    Directory of Open Access Journals (Sweden)

    Steffie H A Brouns

    Full Text Available Emergency department (ED crowding leads to prolonged emergency department length of stay (ED-LOS and adverse patient outcomes. No uniform definition of ED crowding exists. Several scores have been developed to quantify ED crowding; the best known is the Emergency Department Work Index (EDWIN. Research on the EDWIN is often applied to limited settings and conducted over a short period of time.To explore whether the EDWIN as a measure can track occupancy at a Dutch ED over the course of one year and to identify fluctuations in ED occupancy per hour, day, and month. Secondary objective is to investigate the discriminatory value of the EDWIN in detecting crowding, as compared with the occupancy rate and prolonged ED-LOS.A retrospective cohort study of all ED visits during the period from September 2010 to August 2011 was performed in one hospital in the Netherlands. The EDWIN incorporates the number of patients per triage level, physicians, treatment beds and admitted patients to quantify ED crowding. The EDWIN was adjusted to emergency care in the Netherlands: modified EDWIN (mEDWIN. ED crowding was defined as the 75th percentile of mEDWIN per hour, which was ≥0.28.In total, 28,220 ED visits were included in the analysis. The median mEDWIN per hour was 0.15 (Interquartile range (IQR 0.05-0.28; median mEDWIN per patient was 0.25 (IQR 0.15-0.39. The EDWIN was higher on Wednesday (0.16 than on other days (0.14-0.16, p<0.001, and a peak in both mEDWIN (0.30-0.33 and ED crowding (52.9-63.4% was found between 13:00-18:00 h. A comparison of the mEDWIN with the occupancy rate revealed an area under the curve (AUC of 0.86 (95%CI 0.85-0.87. The AUC of mEDWIN compared with a prolonged ED-LOS (≥4 hours was 0.50 (95%CI 0.40-0.60.The mEDWIN was applicable at a Dutch ED. The mEDWIN was able to identify fluctuations in ED occupancy. In addition, the mEDWIN had high discriminatory power for identification of a busy ED, when compared with the occupancy rate.

  9. Obituary: Edwin E. Salpeter (1924-2008)

    Science.gov (United States)

    Trimble, Virginia; Terzian, Yervant

    2009-12-01

    Edwin E. Salpeter, who died 26 November 2008 at his home in Ithaca, NY, belonged to the "second wave" of Jewish scientific refugees from Nazi-dominated Europe, those who left as children just before the onset of WWII and so completed their educations elsewhere. Salpeter was born in Vienna on 3 December 1924, and arrived with his family in Australia in 1939, his father was a physicist and a close friend of Erwin Schrodinger. In Australia, he finished high school, and he entered the University of Sydney at the early age of 16. He received his BS and MSc degrees in physics and mathematics from the University of Sydney, before moving on to a PhD from the University of Birmingham in 1948, for work with Rudolf Peierls on the electrodynamic self-energy of the electron, the first of more than 380 inventoried publications. He had chosen Birmingham over Cambridge or Oxford because of Peierls, and then chose Cornell over Princeton because of Hans Bethe's presence there. His autobiography describes those as two of his very best decisions ever. Marrying psychobiology student Miriam (Mika) Mark less than a year after arriving at Cornell was surely the third, and they remained in Ithaca the rest of their lives, eventually collaborating on some projects in neurobiology before her death in 2000. Their household was a secular one, but (Ed told a colleague) their two daughters received a basic Jewish education "just in case." Daughter Shelley Salpeter and her son Nicholas Buckley were also collaborators with Salpeter on 21st century projects in meta-analysis, epidemiology, and other statistics-heavy problems in biomedicine. Ed Salpeter is survived by his second wife, Antonia (Lhamo) Shouse. Astronomers may be interested to learn that the Cornell press release announcing his death was prepared by Lauren Gold, daughter of Thomas Gold (and Carrie Gold) the co-author of the steady state theory. Apparently, Ed's father Jakob Salpeter late in life considered the anisotropy reported in the

  10. Research in nuclear reactor theory and experimental reactors; Istrazivanja u teoriji nuklearnih reaktora i ekspeimentalni reaktori

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Elektrotehnicki fakultet, Beograd (Yugoslavia)

    1978-05-15

    The paper is devoted to the possibilities of using experimental reactors for scientific research in nuclear power with a stress on problems in nuclear reactor theory. The stationary and nonstationary neutron fields, burnup prediction and analyses as well as fuel element development and the corresponding role of test-reactors were dealt with. It was shown that the investigations in nuclear reactor theory in Yugoslavia were developing continuously and in a useful interaction with experiments on research reactors. The needs for continuing the work on fundamental problems in neutron transport theory and on improving the calculation methods for thermal power reactors, together with the improvement of performances of existing research systems, were pointed out. A new quality in scientific work could be obtained dealing with the problems connected to a possible introduction of test-reactors, and fast systems later on. It was also pleaded for the corresponding orientations in fundamental sciences. (author) Rad je posvecen mogucnostima koriscenja eksperimentalnih reaktora za naucna istrazivanja u nuklearnoj energetici, sa akcentom na probleme teorije nuklearnih reaktora. Obradjena su stacionarna i nestacionarna neutronska polja, predikcija i analize sagorevanja, kao i razvoj gorivnih elemenata te uloga test-reaktora u osvajanju njihove tehnologije. Pokazano je da su se istrazivanja u teoriji nuklearnih reaktora u nas odvijala kontinualno i u korisnoj interakciji sa eksperimentima na istrazivackim reaktorima. Istaknuta je potreba nastavljanja rada na fundamentalnim problemima transportne teorije neutrona i na usavrsavanju metoda proracuna termalnih enerrgetskih reaktora, uz poboljsanje performansi postojecih istrazivackih sistema. Novi kvalitet u naucnom radu bi predstavljala orijentacija na probleme vezane sa eventualnim uvodjenjem test-reaktora, a zatim i brzih sistema. Pledirano je i za odgovarajuca usmeravanja u fundamentalnim naukama. (author)

  11. I. Reactor safety (including comments on criticisms of WASH-1400)

    International Nuclear Information System (INIS)

    1976-01-01

    A major concern in any nuclear power programme is a reactor accident resulting in a large release of radioactivity to the environment. Serious reactor accidents are possible and the risk of such accidents cannot be reduced to zero i.e. absolute safety cannot be assured. All that can be expected is that the measures used to ensure safety in the design and operation of a reactor are such that the risk of accident is reduced to acceptably low levels. No member of the general public is known to have died or been injured as a result of an accident in over 1000 commercial nuclear power reactor-years. Some accidents in power reactors in operation today have come close enough to an environmental release of radioactivity to cause serious public concern about future safety. Apparent inadequacies in safety practices disclosed by former members of the nuclear power industry have added to this concern. To obtain an objective appraisal of the reactor safety issue this report examines the measures taken in the design and operation of nuclear reactors to reduce the probability of accident to acceptably low levels

  12. C-Reactor I and E loading instability limits

    Energy Technology Data Exchange (ETDEWEB)

    Hess, K.W.

    1957-01-24

    The pilot charging of I & E fuel elements has been implemented at C-Reactor under Production Test IP-19-A. It was necessary to provide adequate tube protection against flow interruption by establishing proper trip setting on the Panellit pressure gauges. the administration of these Panellit trip settings is done by trip-before- boiling tube outlet temperature limits, which are similar in principle to the current instability limits. Trip-before-boiling limits for C-Reactor I & E fuel elements loadings are presented in this document.

  13. Magushapud riigihanked / Fredy-Edwin Esse

    Index Scriptorium Estoniae

    Esse, Fredy-Edwin

    2010-01-01

    19. novembril toimunud Nordecon International AS-i aktsionäride erakorralisel koosolekul kiideti heaks tütarfirmade liitmine emaettevõttega uue ärinime Nordecon AS all. Ettevõtte juht Jaano Vink ja selle omanik Toomas Luman leiavad, et edaspidise ebakompetentsuse kõrvaldamiseks ja vahendite kokkuhoidmiseks tuleks riigihanked ühise organisatsiooni alla tuua. Graafik

  14. Maintaining Masculinity in Mid-Twentieth-Century American Psychology: Edwin Boring, Scientific Eminence, and the "Woman Problem".

    Science.gov (United States)

    Rutherford, Alexandra

    2015-01-01

    Using mid-twentieth-century American psychology as my focus, I explore how scientific psychology was constructed as a distinctly masculine enterprise and was navigated by those who did not conform easily to this masculine ideal. I show how women emerged as problems for science through the vigorous gatekeeping activities and personal and professional writings of disciplinary figurehead Edwin G. Boring. I trace Boring's intellectual and professional socialization into masculine science and his efforts to understand women's apparent lack of scientific eminence, efforts that were clearly undergirded by preexisting and widely shared assumptions about men's and women's capacities and preferences.

  15. RA reactor building and installations; Zgrada 'RA' i instalacije

    Energy Technology Data Exchange (ETDEWEB)

    Badrljica, R; Sanovic, V; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1985-08-15

    RA reactor building is made of reinforced concrete and bricks. It is closed facility with a limited number of controlled openings, doors and windows. The site of the building is 100 m above the sea level, 20 m above the mean Danube level and 8 m above the level of the neighbouring stream Mlaka. The building consists of three parts: central prismatic part, annex - surrounding the central part and the sanitary corridor. The biggest space is the reactor hall. In addition to the detailed description and drawings of the reactor building this documents includes design specifications of: electrical installation, water supply system, sewage system, ventilation and heating, gas and compressed air systems. A separate chapter is devoted to fire protection. Zgrada reaktora RA izgradjena je od armiranog betona i opeke, kao zatvoreni objekat ogranicenog broja kontolisanih otvora, sa ogranicenim brojem vrata i prozora. Plato na kojem je zgrada izgradjena nalazi se na 100 m nadmorske visine, na 20 m iznad srednjeg vodostaja Dunava i 8 m iznad nivoa obliznjeg potoka Mlaka. Zgrada se sastoji iz tri dela: sredisnjeg prizmaticnog dela, aneksa - prstenastog okvira sredisnog dela i sanitarnog propusnika. Pojedinacno najveci prostor zauzima reaktorska hala. Pored detaljnog opisa i plana zgrade, ovaj dokument sadrzi projekat elektricne instalacije, projekat vodovoda i kanalizacije, ventilacije i grejanja, instalacije gasa i komprimovanog vazduha. Posebno poglavlje posveceno je protivpozarnoj zastiti.

  16. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1984-12-01

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

  17. Design study of 'HIBLIC-I' reactor cavity

    International Nuclear Information System (INIS)

    Fujiie, Y.

    1984-01-01

    A preliminary conceptual design of a reactor cavity for HIBLIC-1, a heavy ion fusion reactor system, was carried out. Design efforts have been concentrated mainly on the feasibility study of the physical scenario adopted and also on the system integration of the structures and components into a compact reactor cavity. The design features of the reactor are a compact reactor cavity, maximum coolant temperature up to 500 deg C, the protection of the sacrificial wall and cavity wall from radiation, the protection of the sacrificial wall from the pressure transient due to rapid heating, the selection of a ferritic steel HT-9 as the structural material and impurity control, and tritium breeding and recovery. The purpose of this paper is to describe the outline of the reactor cavity design of HIBLIC-1. The objectives of the preliminary conceptual design were to propose the idea and concept in order to constitute the physical scenario without contradiction and to find out the critical and fundamental problems to be studied in future. The cavity configuration and dynamics, tritium breeding and radiation damage, the behavior of a structural material in liquid lithium and tritium recovery are reported. (Kako, I.)

  18. Heroism and Imperialism in the Arctic: Edwin Landseer’s Man Proposes – God Disposes

    Directory of Open Access Journals (Sweden)

    Ingeborg Høvik

    2008-02-01

    Full Text Available Edwin Landseer contributed the painting Man Proposes - God Disposes (Royal Holloway College, Egham, showing two polar bears amongst the remnants of a failed Arctic expedition, to the Royal Academy's annual exhibition of 1864. As contemporary nineteenth-century reviews of this exhibition show, the British public commonly associated Landseer's painting with the lost Arctic expedition of sir John Franklin, who had set out to find the Northwest Passage in 1845. Despite Landseer's gloomy representation of a present-day human disaster and, in effect, of British exploration in the Arctic, the painting became a public success upon its first showing. I will argue that a major reason why the painting became a success, was because Landseer's version of the Franklin expedition's fate offered a closure to the whole Franklin tragedy that corresponded to British nineteenth-century views on heroism and British-ness.

  19. Decommissioning of the ICI TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Parry, D.R.; England, M.R.; Ward, A.; Green, D.

    2000-01-01

    This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

  20. RA Reactor operation and maintenance (I-IX), Part I; Pogon i odrzavanje reaktora RA (I-IX), I Deo

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    The report on RA reactor operation and maintenance for year 1963 is divided in six tasks. This volume contains the introductory report, and three tasks of the final report, namely reactor exploitation, reactivity changes of the RA reactor before repair, planning of refuelling.

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

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

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

  2. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  3. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

    Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

    1997-12-31

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

  4. History of fast reactor development in U.S.A.-I

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sasao, Nobuyki

    2007-01-01

    History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Upgrade of VR-1 training reactor I and C

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Chab, V.

    2003-01-01

    The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-12-01

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

  10. Aurora Borealis, A Painting by Frederic Edwin Church

    Science.gov (United States)

    Love, J. J.

    2015-12-01

    This year marks the sesquicentennial anniversary of the end of the American Civil War. In 1865, the same year as the War's end, the great American landscape artist, Frederic Edwin Church, unveiled Aurora Borealis, a painting that depicts a fantastic, far-northern place, an auroral arch stretched across a quiet night-time sky, above dark mountains and a frozen sea. Church was born in Connecticut, lived in New York, and traveled to Labrador; he would have often seen the northern lights. Church might have also been influenced by the spectacular displays of aurora that were caused by some unusually intense magnetic storms in 1859. Aurora Borealis can certainly be interpreted in terms of 19th-century romanticism, scientific philosophy, and Arctic missions of exploration, all subjects of interest to Church. As with so many of his paintings, Church's meticulous attention to detail in Aurora Borealis reveals his deep admiration of nature. But his depiction of auroral light is a curious and possibly intentional departure from natural verisimilitude. Some art historians have suggested that Church painted Aurora Borealis as a subdued tribute to the end of the Civil War, with the drapery of auroral light forming an abstract representation of the American flag. If so, then colors of the flag have been unfurled across a cold and barren landscape, not in extravagant celebration, but in somber recognition of the reality of post-war desolation and an uncertain future.

  11. Edwin Grant Dexter: an early researcher in human behavioral biometeorology

    Science.gov (United States)

    Stewart, Alan E.

    2015-06-01

    Edwin Grant Dexter (1868-1938) was one of the first researchers to study empirically the effects of specific weather conditions on human behavior. Dexter (1904) published his findings in a book, Weather influences. The author's purposes in this article were to (1) describe briefly Dexter's professional life and examine the historical contexts and motivations that led Dexter to conduct some of the first empirical behavioral biometeorological studies of the time, (2) describe the methods Dexter used to examine weather-behavior relationships and briefly characterize the results that he reported in Weather influences, and (3) provide a historical analysis of Dexter's work and assess its significance for human behavioral biometeorology. Dexter's Weather influences, while demonstrating an exemplary approach to weather, health, and behavior relationships, came at the end of a long era of such studies, as health, social, and meteorological sciences were turning to different paradigms to advance their fields. For these reasons, Dexter's approach and contributions may not have been fully recognized at the time and are, consequently, worthy of consideration by contemporary biometeorologists.

  12. WITAMIR-I: A tandem mirror power reactor

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Beyer, J.B.

    1983-01-01

    A conceptual design of a near term commercial tandem mirror power reactor will be presented. The basic configuration utilizes Yin-yang minimum B end plugs with inboard thermal barriers, which are pumped by neutral beam injection. The maximum magnetic fields are 6.1 T, 8.1 T and 15 T in the central cell, Yin-yang, and thermal barrier magnets, respectively. The blanket utilizes Pb 83 Li 17 as the coolant and breeder, and HT-9 as the structural material. This configuration yields a high energy multiplication (1.37), a sufficient tritium breeding ratio (1.07) and has a major advantage with respect to maintenance. A single stage direct convertor is used at one end and an electron thermal dump at the other end. The plasma Q is 28 at a fusion power level of 3000 MWsub(th); the net electrical output is 1530 MWe and the overall efficiency is 39%. Cost estimates indicate that WITAMIR-I is competitive with recent tokamak power reactor designs. (author)

  13. RA research reactor - properties and experimental capabilities; Istrazivacki reaktor RA - Tehnicke karakteristike i eksploatacione mogucnosti

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described. Prikazani su fragmenti iz eksploatacije reaktora kao i stanje opreme, posle 18 godina rada. Na kraju je dat prikaz sta je preduzeto i sta se preduzima da se poboljsaju karakteristike i poveca sigurnost i bezbednost rada za sledecih 15-20 godina.

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

    International Nuclear Information System (INIS)

    Lolic, B.; Jovanovic, S.

    1961-12-01

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

  15. Reactor noise analysis applications in NPP I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

    2006-07-01

    Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

  16. Socio-economic impact of nuclear reactor decommissioning at Vandellos I NPP

    International Nuclear Information System (INIS)

    Liliana Yetta Pandi

    2013-01-01

    Currently nuclear reactors in Indonesia has been outstanding for more than 30 years, the possibility of nuclear reactors will be decommissioned. Closure of the operation or decommissioning of nuclear reactors will have socio-economic impacts. The socioeconomic impacts occur to workers, local communities and wider society. In this paper we report on socio-economic impacts of nuclear reactors decommissioning and lesson learned that can be drawn from the socio-economic impacts decommissioning Vandellos I nuclear power plant in Spain. Socio-economic impact due to decommissioning of nuclear reactor occurs at installation worker, local community and wider community. (author)

  17. Research and development on next generation reactor (phase I)

    International Nuclear Information System (INIS)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)

  18. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  19. Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

    2003-12-01

    Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

  20. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte I

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R; Ahnert, C; Naudin, A E; Martinez Fanegas, R; Minguez, E; Rovira, A

    1976-07-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

  1. Psi Chi/APA Edwin B. Newman Graduate Research Award: Joseph H. Hammer

    Science.gov (United States)

    American Psychologist, 2009

    2009-01-01

    Joseph H. Hammer, recipient of the Psi Chi/APA Edwin B. Newman Graduate Research Award, is cited for an outstanding research paper whose findings provide important evidence regarding the promise of a male-sensitive approach to mental health marketing and empirically support the inclusion of theory-driven enhancements in group-targeted mental…

  2. Practical polyphenolics: from structure to molecular recognition and physiological action, by Edwin Haslam.[Book review

    Science.gov (United States)

    Richard W. Hemingway

    1998-01-01

    Hemingway’s book review brings into focus Edwin Haslam's career, devoted to defining the significance of plant polyphenols. That historical perspective focuses on the progress made in this science over the last 30 years. Most important, this book demonstrates the myriad ways that plant polyphe­nols influence our lives. Professor Haslam makes a strong argument for...

  3. Cynthia J. Najdowski: Psi Chi/APA Edwin B. Newman Graduate Research Award

    Science.gov (United States)

    American Psychologist, 2012

    2012-01-01

    Presents a short biography of the winner of the American Psychological Association's Psi Chi/APA Edwin B. Newman Graduate Research Award. The 2012 winner is Cynthia J. Najdowski for an outstanding research paper that examines how jurors' judgments are influenced by a juvenile defendant's confession and status as intellectually disabled. Through…

  4. Microchannel Reactors for ISRU Applications Using Nanofabricated Catalysts, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Makel Engineering, Inc. (MEI) and USRA propose to develop microchannel reactors for In-Situ Resources Utilization (ISRU) using nanofabricated catalysts. The proposed...

  5. RA reactor operation and maintenance; Pogon i odrzavanje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air.

  6. Reactor inventory monitoring system for Angra-1 reactor; Sistema de monitoracao de inventario do reator para usina nuclear Angra I

    Energy Technology Data Exchange (ETDEWEB)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M. [Furnas Centrais Eletricas S.A., Rio de Janeiro, RJ (Brazil); Soares, Milton [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Lab. de Monitoracao de Processos

    1996-07-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  7. Measuring temperature coefficient of TRIGA MARK I reactor by noise analysis

    International Nuclear Information System (INIS)

    Soares, P.A.

    1975-01-01

    The transfer function of TRIGA MARK I Reactor is measured at power zero (5w) and power 118Kw, in the frequency range of 0.02 to 0.5 rd/s. The method of intercorrelation between a pseudostochasticbinary signal is used. A simple dynamic model of the reactor is developed and the coefficient of temperature is estimated [pt

  8. Photocatalytic reactors for treating water pollution with solar illumination. I: a simplified analysis for batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

    2003-07-01

    Usual applications of photocatalytic reactors for treating wastewater exhibit the difficulty of handling fluids having varying composition and/or concentrations; thus, a detailed kinetic representation may not be possible. When the catalyst activation is obtained employing solar illumination an additional complexity always coexists: solar fluxes are permanently changing with time. For comparing different reacting systems under similar operating conditions and to provide approximate estimations for scaling up purposes, simplified models may be useful. For these approximations the model parameters should be restricted as much as possible to initial physical and boundary conditions such as: initial concentrations (expressed as such or as TOC measurements), flow rate or reactor volume, irradiated reactor area, incident radiation fluxes and a fairly simple experimental observation such as the photonic efficiency. A combination of a new concept: the ''actual observed photonic efficiency'' with ideal reactor models and empirical kinetic rate expressions can be used to provide rather simple working equations that can be efficiently used to describe the performance of practical reactors. In this paper, the method has been developed for the case of a photocatalytic batch reactor (PBR). (orig.)

  9. Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

    International Nuclear Information System (INIS)

    Bultman, J.

    1992-04-01

    Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

  10. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    Brynda, W.J.; Powell, R.W.

    1978-08-01

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  11. Hydrologic assessment of the Edwin B. Forsythe National Wildlife Refuge

    Science.gov (United States)

    Wieben, Christine M.; Chepiga, Mary M.

    2018-03-19

    The Edwin B. Forsythe National Wildlife Refuge (hereafter Forsythe refuge or the refuge) is situated along the central New Jersey coast and provides a mixture of freshwater and saltwater habitats for numerous bird, wildlife, and plant species. Little data and information were previously available regarding the freshwater dynamics that support the refuge’s ecosystems. In cooperation with the U.S. Fish and Wildlife Service, the U.S. Geological Survey conducted an assessment of the hydrologic resources and processes in the refuge and surrounding areas to provide baseline information for evaluating restoration projects and future changes in the hydrologic system associated with climate change and other anthropogenic stressors.During spring 2015, water levels were measured at groundwater and surface-water sites in and near the Forsythe refuge. These water-level measurements, along with surface-water elevations obtained from digital elevation models, were used to construct water-table-elevation and depth-to-water maps of the refuge and surrounding areas. Water-table elevations in the refuge ranged from sea level to approximately 65 feet above sea level; in most of the refuge, the water-table elevation was within 3 feet of sea level. The water-table-elevation map indicates that the direction of shallow groundwater flow at the regional scale is generally from west to east (much of it from the northwest to the southeast), and groundwater moves downgradient from the uplands toward major groundwater discharge areas consisting of coastal streams and wetlands. The depth to water is estimated to be less than 2 feet for approximately 86 percent of the refuge, which coincides closely with the percentage of wetland area in the refuge. Depth to water in excess of 20 feet below land surface is limited to higher elevation areas of the refuge.Streamflow data collected at continuous-record streamgages and partial-record stations within the Mullica-Toms Basin were summarized

  12. Possibilities for power reactor structural material and fuel testing in reactor RA; Mogucnosti reaktora RA za testiranje konstrukcionih materijala i goriva energetskih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Lazarevic, Dj; Stefanovic, D; Cupac, S; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    Nuclear reactor RA at Vinca has been designed as a high flux general purpose research reactor. Among other it was intended to play a role of material testing reactor. A scope of activities of Material Laboratory and Reactor RA Department of Boris Kidric Institute is presented in this report. Reactor RA capacity for reactor structural material and fuel irradiation is also described. The increase of RA reactor irradiation capacity is based on the improvement of VISA type fuel channel for fast neutron irradiations, as well as on the general neutron flux increase, due to introduction of highly enriched uranium fuel into reactor core and the advanced in-core fuel management. The irradiation capacities described allow for the reactor material and fuel testing to the considerable extent. Istrazivacki reaktor RA u Vinci je projektovan kao visokofluksni istrazivacki reaktor opste namene. Pored ostalog, on je namenjen i za testiranje reaktorskih konstrukcionih materijala i goriva. U radu je dat pregled aktivnosti Laboratorije za materijale IBK i reaktora RA na tom podrucju, kao i opis povecanih mogucnosti reaktora RA za ozracivanje reaktorskih materijala i goriva u cilju njihovog testiranja. Povecanje mogucnosti reaktora RA zasniva se na usavrsavanju specijalnog gorivnog kanala tipa VISA (za ozracivanje materijala brzim neutronima), kao i na opstem povecanju neutronskog fluksa na osnovu uvodjenja i nacina koriscenja visokoobogacenog uranskog goriva u reaktoru RA. Opisane mogucnosti reaktora RA dozvoljavaju u znatnoj meri ispitivanje konstrukcionih materijala i goriva energetskih reaktora.

  13. Reactor safeguards system assessment and design. Volume I

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.; Bennett, H.A.; Hulme, B.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore, should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems

  14. Materials data base for fusion reactors-I

    International Nuclear Information System (INIS)

    Iwata, S.; Nogami, A.; Ishino, S.; Mishima, Y.; Takao, Y.; Aruga, T.; Shiraishi, K.

    1982-01-01

    The materials data base is a set of experimental and/or calculated data being compiled to meet the broad needs for materials data by taking advantage of the data base management systems. In this paper the objective of such computerized data base is described and the characteristics of fusion reactor materials are discussed from the viewpoint of the data base development. The near-term emphasis of the development has been put on the irradiation data for 316 type stainless steels. Through the test of this small data base, it can be concluded that this approach is promising for materials data base management and for the establishment of the interface between fusion reactor designer and materials investigator. (orig.)

  15. Structural integrity analysis of reactor coolant pump flywheel(I)

    International Nuclear Information System (INIS)

    Kim, Young Jin

    1986-01-01

    A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

  16. Michael K. Scullin: Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2011-11-01

    Presents Michael K. Scullin as the 2011 winner of the American Psychological Association Psi Chi/APA Edwin B. Newman Graduate Research Award. "For an outstanding research paper that examines the relationship between prospective memory in executing a goal and various lapses of time from 20 minutes up to a 12- hour wake delay and a 12-hour sleep delay. The results suggest that consolidation processes active during sleep increase the probability of goal execution. The paper, titled 'Remembering to Execute a Goal: Sleep On It!' was published in Psychological Science in 2010 and was the basis for Michael K. Scullin's selection as the recipient of the 2011 Psi Chi/APA Edwin B. Newman Graduate Research Award. Mark A. McDaniel, PhD, served as faculty research advisor." (PsycINFO Database Record (c) 2011 APA, all rights reserved). 2011 APA, all rights reserved

  17. Physics of Fast and Intermediate Reactors. V. I. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. V. I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1962-03-15

    in all cases that of heir presentation during the Seminar. Changes have been made where it was considered that these would enhance the usefulness of these volumes as reference books. The subject grouping adopted is given below. Volume I - I. Neutron Physics: I.1. Data requirements, I.2. Cross-section measurements, I.3. Fission properties, I.4. Nuclear theory, I.5. Multi-group cross-sections; II. Integral Experiments: II.1. Critical experiments, II.2. Other integral experiments, II.3. Theoretical correlations; Volume II - III. Reactor Theory: III.1. Calculation methods, III.2. Effects of cross-section errors, III.3. Reactivity effects, III.4. Long-term effects, III.5. Reactor concept studies; Volume III - IV. Reactor Dynamics: IV.1. Kinetics, IV.2. Stability, IV.3. Doppler effect, IV.4. Safety problems; V. Physics of Specific Reactors.

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

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Raickovic, N.; Radivojevic, J.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1983-12-01

    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 [sr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-15

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

  20. RA reactor operation and maintenance in 1992, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktoro RA u 1992. Godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems. [Serbo-Croat] U toku 1992 godine poslovi u okviru projekta 'Istrazivacki nuklearni reaktor RA' obavljani su u skladu sa programom i planom rada. Osnovne aktivnosti na kojima je radjeno odnosile su se na revitalizaciju reaktora RA, kao i na odrzavanje opreme. U ovom periodu reaktor nije bio u pogonu. Svo osoblje je bilo angazovano na poslovima rekonstrukcije i modernizacije postojecih i dogradnje novih reaktorskih sistema, na odrzavanju opreme a deo tehnickog osoblja je bio obucavan za vrsenje odgovarajucih poslova u pogonu i odrzavanju opreme.

  1. Neutronics and thermohydraulics of the reactor C.E.N.E.-Part I

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Naudin, A. E.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    In this report the analysis of neutronics (both statics and kinetics), of the 10 MWt swimming pool reactor C.E.N.E, is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking, carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-07-01

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

  3. An analysis of power transients observed in SPERT I reactors

    International Nuclear Information System (INIS)

    Clancy, B.E.; Connolly, J.W.; Harrington, B.V.

    1976-04-01

    The analytical method described in Part I of this series has been applied to the calculation of spert I transients performed with higher initial moderator temperatures and also to those performed in a highly undermoderated core. Reasonable agreement has been obtained between calculated and experimental burst data. (author)

  4. Revised Revised Edwin Etieyibo The Ethics of Government ...

    African Journals Online (AJOL)

    Jimmy Gitonga

    enterprises (SOEs) by the Federal Government of Nigeria (FGN) is ethical. ... ownership of an enterprise, business or agency is transferred from the public sector ... I begin by examining some issues that arise from the idea of privatization. Next ...

  5. Study on Optimization of I and C Architecture for Research Reactors Using Bayesian Networks

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khaili Ur; Shin, Jinsoo; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-07-01

    The optimization in terms of redundancy of modules and components in Instrumentation and Control (I and C) architecture is based on cost and availability assuming regulatory requirements are satisfied. The motive of this study is to find an optimized I and C architecture, either in hybrid formation, fully digital or analog, with respect to system availability and relative cost of architecture. The cost of research reactors I and C systems is prone to have effect on marketing competitiveness. As a demonstrative example, the reactor protection system of research reactors is selected. The four cases with different architecture formation were developed with single and double redundancy of bi-stable modules, coincidence processor module, and safety or protection circuit actuation logic. The architecture configurations are transformed to reliability block diagram (RBD) based on logical operation and function of modules. A Bayesian Network (BN) model is constructed from RBD to assess availability. The cost estimation was proposed and reliability cost index RI was suggested.

  6. Study on Optimization of I and C Architecture for Research Reactors Using Bayesian Networks

    International Nuclear Information System (INIS)

    Rahman, Khaili Ur; Shin, Jinsoo; Heo, Gyunyoung

    2013-01-01

    The optimization in terms of redundancy of modules and components in Instrumentation and Control (I and C) architecture is based on cost and availability assuming regulatory requirements are satisfied. The motive of this study is to find an optimized I and C architecture, either in hybrid formation, fully digital or analog, with respect to system availability and relative cost of architecture. The cost of research reactors I and C systems is prone to have effect on marketing competitiveness. As a demonstrative example, the reactor protection system of research reactors is selected. The four cases with different architecture formation were developed with single and double redundancy of bi-stable modules, coincidence processor module, and safety or protection circuit actuation logic. The architecture configurations are transformed to reliability block diagram (RBD) based on logical operation and function of modules. A Bayesian Network (BN) model is constructed from RBD to assess availability. The cost estimation was proposed and reliability cost index RI was suggested

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

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1964-12-01

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

  8. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  9. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters

  10. Sodium fast reactor safety and licensing research plan - Volume I

    International Nuclear Information System (INIS)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-01-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  11. Super Bowli teooria soovitab aktsiaid osta / Fredy-Edwin Esse

    Index Scriptorium Estoniae

    Esse, Fredy-Edwin

    2011-01-01

    Super Bowli teooria kohaselt tõuseb Dow Jones, kui võidab algupärase NFL-i meeskond. MarketWatchi peaökonomist Irwin Keller avaldab arvamust, miks tasuks investoritel Super Bowli tulemustele tähelepanu pöörata

  12. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  13. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  14. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  15. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

    International Nuclear Information System (INIS)

    Raisic, N.

    1962-09-01

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. I and C system at TRIGA - ICN reactor after more than 20 years of operation

    International Nuclear Information System (INIS)

    Ionila, M.; Preda, M.

    2002-01-01

    An I and C system that is involved in a nuclear safety function has to be itself safe in operation, strictly performing the survey of those parameters, which are linked to the safety function. The precision of such a system is sufficient for the safety function, but for a more accurate evaluation of the in-core experimental phenomena, the presence of a data acquisition and processing system is needed. The two systems must be together taken into account by the reactor operation. The data acquisition and processing system designated for the monitoring of the stationary or the slow-varying processes allow the safety function evaluation from the point of view of the statistics of the effective reactor operation time along a certain period of time. The evaluation of the unscheduled reactor shutdowns determined by those reactor systems having safety functions with the percentage contribution of each system is presented. The data were selected from the annual operation reports for the reactor and the reactor installations in the period 1981-1999

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  19. Edwin B. Wilson and the rise of mathematical economics in America, 1920-1940

    OpenAIRE

    Carvajalino, Juan

    2017-01-01

    In the paper, Edwin B. Wilson's influence on the rise of mathematical economics in America between the 1920s and 1940s is explored. The focus is laid on showing how on the grounds of his foundational ideas about science Wilson worked at the organizational and educational fronts to modernize economics, at this at three levels. First, the paper shows the ways in which around 1930 Wilson was key, at the nationwide level, in the constitution of the first organized community of American mathematic...

  20. Connor H. G. Patros: Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2015-11-01

    The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and APA. The award was established to recognize young researchers at the beginning of their professional lives and to commemorate both the 50th anniversary of Psi Chi and the 100th anniversary of psychology as a science (dating from the founding of Wundt's laboratory). The 2015 recipient is Connor H. G Patros. Patros was chosen for "an excellent research paper that examines the complex relationship between working memory, choice-impulsivity, and the attention-deficit/hyperactivity disorder (ADHD) phenotype." Patros's award citation, biography, and a selected bibliography are presented here. (c) 2015 APA, all rights reserved).

  1. APA/Psi Chi Edwin B. Newman Graduate Research Award: Samantha F. Anderson.

    Science.gov (United States)

    2017-12-01

    The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and the American Psychological Association. The award was established to recognize young researchers at the beginning of their professional lives and to commemorate both the 50th anniversary of Psi Chi and the 100th anniversary of psychology as a science (dating from the founding of Wundt's laboratory). The 2017 recipient is Samantha F. Anderson, who was chosen for "an exceptional research paper that responds to psychology's 'replication crisis' by outlining a broader view of success in replication." Her award citation, biography, and a selected bibliography are presented here. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  2. Edwin G. Boring: The Historian's Path in the Pages of The American Journal of Psychology.

    Science.gov (United States)

    Gallagher, Shawn R

    2017-01-01

    Although he is best known for his classic textbook, A History of Experimental Psychology, Edwin Garrigues Boring published dozens of articles in The American Journal of Psychology and used its various formats to guide the discipline in the early 20th century. This report reviews a small sample of his publications, including obituaries, notes, and experimental articles, and presents them in historical and biographical context. A central objective is to show how Boring shared the values of his structuralist training with the emerging American schools and how time allowed him to reconsider his approach to history and the legacy of his iconic mentor, Edward Bradford Titchener.

  3. Operation and maintenance of the RA nuclear reactor for 1977, Report Annex I; Rad i iskoriscenost reaktora RA u 1977. godini, izvestaj Prilog I

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    RA reactor operation plan was fulfilled, meaning 28583 MWh. In addition to 183 days of operation at full power, during 1977 the reactor was operated for 14 days at lower power level according to the demand of users. The utilization level of rector irradiation capability (neutron flux and time of operation) was 14%. This annex includes detailed statistical data about reactor operation, utilization, power level, savings concerned with new 80% enriched fuel. All the 9 vertical experimental channels were used for irradiation in the reactor core. Digression from the action plan were caused by refueling and demand od the users. There have been 11 safety shutdowns, of which 6 caused by power cuts, 4 due to failure of the instruments, and 1 due to earthquake in March 1977. [Serbo-Croat] Planirani rad reaktora na nominalnoj snazi ostvaren je u iznosu od 28583 MWh. U toku 1977. godine reaktor je radio 14 dana na manjim snagama i u posebnom eksperimentalnom rezimu na zahtev korisnika. Iskoriscenost kapaciteta reaktora za ozracivanje uzoraka (na bazi neutronskog fluksa i vremena rada reaktora) bila je 41%. Ovaj izvestaj sadrzi detaljne statisticke podatke o radu i iskoriscenosti reaktora, podatke o ustedi goriva prelaskom na 80% obogaceno gorivo. Korisceno je svih 9 vertikalnih eksperimentalnih kanala u aktivnoj zoni. Uzroci odstupanja od plana rada osim zamene goriva bili su zahtevi korisnika. Bilo je 11 sigurnosnih zaustavljanja, 6 usled nestanka napona, 4 usled kvarova opreme i instrumentacije, i 1 put usled zemljotresa u martu.

  4. Investigations on the retention of 131I by an iodine filter of a pressurized water reactor

    International Nuclear Information System (INIS)

    Deuber, H.; Gerlach, K.

    1983-09-01

    The retention of 131 I by an equipment room exhaust filter of a German pressurized water reactor was determined by various methods to particularly obtain reliable results. Moreover, investigations were performed to clarify the reason for aging of the carbon contained in the iodine filter mentioned. The actual retention of the organic 131 I, corresponding to a value of 99.9%, was limited by 131 I in the form of penetrating iodine compounds. It was lower than the retention of CH 3 131 I under layout conditions by more than one order of magnitude. The aging was essentially caused by the adsorption of low-volatile organic compounds. (orig.) [de

  5. Investigations on the retention of I-131 by an iodine filter of a pressurized water reactor

    International Nuclear Information System (INIS)

    Deuber, H.; Gerlach, K.

    1984-01-01

    The retention of I-131 by an equipment room exhaust filter of a German pressurized water reactor was determined by various methods to particularly obtain reliable results. Moreover, investigations were performed to clarify the reason for aging of the carbon contained in the iodine filter mentioned. The actual retention of the organic I-131, corresponding to a value of 99.9%, was limited by I-131 in the form of penetrating iodine compounds. It was lower than the retention of CH 3 I-131 under layout conditions by more than one order of magnitude. The aging was essentially caused by the adsorption of low-volatile organic compounds. (orig.) [de

  6. Maintainability considerations for the central cell in WITAMIR-I, a conceptual design of a tandem mirror fusion power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.

    1980-10-01

    The concepts for maintaining the central cell reactor components for WITAMIR-I are described. WITAMIR-I is a conceptual tandem mirror fusion power reactor utilizing thermal barriers designed by the University of Wisconsin-Madison. Unique solutions to the difficult problems of routine blanket replacement and maintenance are proposed. Solutions are also proposed for maintaining the central cell coils and the shield

  7. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1984. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-12-15

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

  9. Effect of fallout on measurement of 131I around nuclear reactors

    International Nuclear Information System (INIS)

    Paperiello, C.J.; Matuszek, J.M.

    1976-01-01

    In early July 1974, 131 I produced by detonaion of a nuclear device by the People's Republic of China appeared in fallout over New York State. Radioiodine levels in milk were measured using a β-γ coincidence system with a sensitivity of 0.02 pCi/liter. Peak levels of 1.6 pCi/liter of milk in early July tapered off to approx.0.1 pCi/liter by early October. When fresh pasture growth ceased and supplemental feed was provided, radioiodine was no longer detectable. This episode shows that operators of light-water power reactors must analyze background samples collected some distance from the reactor site to meet the present U.S. Nuclear Regulatory Commission analytical requirements for 131 I as expressed in Appendix I to 10CFR50 and Regulatory Guide 1.42

  10. Obituary: Raymond Edwin White Jr., 1933-2004

    Science.gov (United States)

    Liebert, James William

    2004-12-01

    traditions. Earlier in 1971-72, Ray served as Program Officer for Stars and Stellar Evolution in the Astronomy Section of the National Science Foundation. Ray was one of the three "originators" of "The Inspiration of Astronomical Phenomena" (INSAP) Conferences. These conferences provide scholarly discussions on the many and variegated cultural impacts of the perceptions about the day- and night-time sky, thus providing a forum for a broad sampling of artists, historians, philosophers, and scientists to get together, compare notes, and ask questions of one another. The INSAP Conferences have taken place near Castel Gandolfo Italy, on the island of Malta, near Palermo Italy, and at Oxford University in England. Ray's scholarship also was manifest in his activities as editor. For some years in the 1990s, he edited two astronomy journals, The Astronomy Quarterly and Vistas in Astronomy. Raymond E. White, Jr., is survived by his wife Ruby E. (nee Fisk), his high school sweetheart at Heidelberg High in Germany. Their children include Raymond E. White III (Professor of Physics and Astronomy at the University of Alabama, Tuscaloosa), Kathleen M. (White) Wade, and Kevin D. White. Ray was proud of two beautiful granddaughters, Charlotte R. Wade and Sarah E. Wade. Ray was proud of his early role with Steward Observatory Director Bart Bok in the commissioning of the "90-inch" reflector at the University of Arizona site on Kitt Peak in 1969. He built the direct camera, and was invited by his close friend Bok to share the "first light" of this telescope, now renamed the Bok 2.3-m telescope. When Professor Bok passed away, the astronomy magazine Sky & Telescope invited Ray to write an article which was entitled "Bart J. Bok (1906-83): Personal Memoir from a Grandson." (Bok mentored Ivan R. King, who was Ray's thesis advisor.) In his concluding remarks, Ray wrote, "The aspect of Bart J. Bok I will miss the most is his exuberance for the art of astronomy." We will also miss greatly this

  11. Development of the reactor antineutrino detection technology within the iDream project

    Science.gov (United States)

    Gromov, M.; Kuznetsov, D.; Murchenko, A.; Novikova, G.; Obinyakov, B.; Oralbaev, A.; Plakitina, K.; Skorokhvatov, M.; Sukhotin, S.; Chepurnov, A.; Etenko, A.

    2017-12-01

    The iDREAM (industrial Detector for reactor antineutrino monitoring) project is aimed at remote monitoring of the operating modes of the atomic reactor on nuclear power plant to ensure a technical support of IAEA non-proliferation safeguards. The detector is a scintillator spectrometer. The sensitive volume (target) is filled with a liquid organic scintillator based on linear alkylbenzene where reactor antineutrinos will be detected via inverse beta-decay reaction. We present first results of laboratory tests after physical launch. The detector was deployed at sea level without background shielding. The number of calibrations with radioactive sources was conducted. All data were obtained by means of a slow control system which was put into operation.

  12. Cynthia J. Najdowski: Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2012-11-01

    Presents a short biography of the winner of the American Psychological Association's Psi Chi/APA Edwin B. Newman Graduate Research Award. The 2012 winner is Cynthia J. Najdowski for an outstanding research paper that examines how jurors' judgments are influenced by a juvenile defendant's confession and status as intellectually disabled. Through the use of a mock trial experiment, the research revealed that jurors discounted a juvenile's coerced confession and sometimes used intellectual disability as a mitigating factor. Attribution theory and the discounting principle were used to identify the psychological mechanisms underlying this effect. The paper, titled 'Understanding Jurors' Judgments in Cases Involving Juvenile Defendants,' was published in Psychology, Public Policy, and Law in October 2011 and was the basis for Najdowski's selection as the recipient of the 2012 Psi Chi/APA Edwin B. Newman Graduate Research Award. Bette L. Bottoms, PhD, served as faculty supervisor. Najdowski's Award citation and a selected bibliography are also presented. PsycINFO Database Record (c) 2012 APA, all rights reserved.

  13. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

  14. Preliminary conceptual design for electrical and I and C system of a new research reactor

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S.

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements

  15. The behavior of 131I in polymetatelluric acid irradiated in the nuclear reactor

    International Nuclear Information System (INIS)

    Teofilovski, C.

    1966-01-01

    Polymetarelluric acid, whose composition is (H 2 TeO 4 ) n , is successfully used at he Institute as a target for obtaining 131 I in the reactor. It is prepared by hearing orthotelluric acid in air at 160 deg C of in a steam of water vapor at 208 deg C. Analysis of the valency states of 131 I in irradiated (H 2 TeO 4 ) n prepared in either of the above ways shows a variable ratio of reduced and oxidized forms. A considerable increase of the reduced forms with increasing integral thermal neutron flux during irradiation in the reactor in the given interval has also been observed. In order to explain the above phenomenon (H 2 TeO 4 ) n was irradiated in the reactor under different conditions, with measurement of the wall temperature of the quartz ampoules containing the target material. Yields of reduced and oxidized form of 131 I were determined immediately after irradiation and after annealing of the target at temperatures from 60 deg C to 150 deg C. A considerable decrease in the yield of the reduced forms of 131 I on target annealing above 100 deg C was observed (author)

  16. The behavior of {sup 131}I in polymetatelluric acid irradiated in the nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teofilovski, C [Institute of Nuclear Sciences Boris Kidric, Hot Laboratory Department, Vinca, Beograd (Serbia and Montenegro)

    1966-01-15

    Polymetarelluric acid, whose composition is (H{sub 2}TeO{sub 4}){sub n}, is successfully used at the Institute as a target for obtaining {sup 131}I in the reactor. It is prepared by hearing orthotelluric acid in air at 160 deg C of in a steam of water vapor at 208 deg C. Analysis of the valency states of {sup 131}I in irradiated (H{sub 2}TeO{sub 4}){sub n} prepared in either of the above ways shows a variable ratio of reduced and oxidized forms. A considerable increase of the reduced forms with increasing integral thermal neutron flux during irradiation in the reactor in the given interval has also been observed. In order to explain the above phenomenon (H{sub 2}TeO{sub 4}){sub n} was irradiated in the reactor under different conditions, with measurement of the wall temperature of the quartz ampoules containing the target material. Yields of reduced and oxidized form of {sup 131}I were determined immediately after irradiation and after annealing of the target at temperatures from 60 deg C to 150 deg C. A considerable decrease in the yield of the reduced forms of {sup 131}I on target annealing above 100 deg C was observed (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  18. Det indskrevne publikum i politiske kommentarer

    DEFF Research Database (Denmark)

    Bengtsson, Mette

    2016-01-01

    I denne artikel fremsættes en kritik af kommentargenren i en dansk kontekst. Med udgangspunkt i 90 kommentarartikler fra landsdækkende aviser under folketingsvalget i 2011 laves en analyse af genrens indskrevne publikum efter Edwin Blacks analyseforskrifter i tre trin: Først analyseres materialet...

  19. The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Miller, R.L.

    1989-01-01

    The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

  20. Conceptual design of laser fusion reactor, SENRI-I - 1. concept and system design

    International Nuclear Information System (INIS)

    Ido, S.; Naki, S.; Norimatsu, T.

    1981-01-01

    Design features of a laser fusion reactor concept SENRI-I and new concepts are reviewed and discussed. The unique feature is the utilization of a magnetic field to guide and control the inner liquid Li flow. Basic requirements and typical parameters used in the design are presented. Items to be discussed are constitution of the system, performance of liquid Li flow, neutronics, thermo-electric cycle, fuel cycle and new concepts

  1. Management and inspection of integrity of spent fuel from IRT MEPhI research reactor

    International Nuclear Information System (INIS)

    Aden, V.G.; Bulkin, S.Y.; Sokolov, A.V.; Bushuev, A.V.; Redkin, A.F.; Portnov, A.A.

    2002-01-01

    The information on wet storage and dry storage of the spent nuclear fuel (SNF) of the IRT MEPhI reactor and experience from SNF shipment for reprocessing are presented. The procedure and a facility for nondestructive inspection of local power density fields and the burnup of fuel assemblies based on studying the γ-activity of some fission products generated in U 235 and procedure for inspection of the fuel element cladding leak tightness are described. (author)

  2. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  3. International Working Group on Fast Reactors Thirteenth Annual Meeting. Summary Report. Part I

    International Nuclear Information System (INIS)

    1980-09-01

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

  4. Regulatory use the classification security systems of I and C in VVER type reactors

    International Nuclear Information System (INIS)

    Ilizastegui Perez, F.

    1998-01-01

    Presently work the author proposes a classification to the system I and C to the VVER 440 type reactor in categories the regulatory control with a view to establishing the degree to the attention that the regulator should pay to these systems, leaving the importance that have the same ones for the security the installation, during the execution the works that are carried out with this equipment in the stages construction, setting in service and exploitation

  5. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    International Nuclear Information System (INIS)

    Khalil, Rah Man; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan

    2012-01-01

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article

  6. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khalil, Rah Man; Heo, Gyun Young [Kyung Hee Univ., Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article.

  7. IPR-RI TRIGA MARK I reactor and the neutron activation analysis at CDTN/CNEN

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Kastner, Geraldo F.; Amaral, Angela M.; Souza, Wagner de; Maretti, Fausto Junior; Leal, Alexandre S.

    2008-01-01

    The IPR-R1 TRIGA Mark I research reactor started up in 1960. It is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. Join to the reactor, the Laboratory for Neutron Activation Analysis has been developing its activities since 1960. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for relevant percentage of CDTN's analysis demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays several elements - Ag, Al, Au, As, Ba, Br, Ca, Cd, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Ga, Hf, Hg, Ho, K, La, Mg, Mn, Mo, Na, Nd, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr - are determined in several matrices and range of concentrations. In Brazil, CDTN is the only Institute that fully masters the instrumental neutron activation analysis k0-method determining short, medium and long half-life radionuclides using its own nuclear reactor. The good performance of the reactor is pointed out in a table with experimental and certified values for Certified Reference Materials. (authors)

  8. Metallurgical and reactor physics aspects of using low enrichment fuel in Safari-I

    International Nuclear Information System (INIS)

    1978-09-01

    The feasibility of using lower than 93% enriched fuel in the SAFARI-I research and materials testing reactor is reviewed. Metallurgical experiments show that, using standard U-Al alloy technology and keeping the 235 U loading per element constant without altering the fuel plate thickness, a maximum of 35 weight percent of uranium in the meat can be achieved. This corresponds to using a minimum enrichment of 40% 235 U in order to retain the same mass of 235 U in the core. Even then a loss of approximately 3,3% in reactivity is calculated, which is more than the 2,8% sup(deltak)/k which is normally allowed for burnup. Using current U-Al alloy fuel technology, and an enrichment of approximately 45% 235 U, no changes in core configuration or coolant requirements will be necessary. The use of 20% enriched uranium will require the development of a new fuel design and technology if drastic redesign and modification of the reactor and coolant circuits is to be avoided. Without such new technology, the redesign and modification of the reactor will cost upwards of 3 million dollars and take up to 5 years to complete, requiring a complete shutdown of the reactor for approximately 2 years

  9. Calibration of new I and C at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, Martin; Jurickova, Monika

    2011-01-01

    The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

  10. Duty or dream? Edwin G. Conklin's critique of eugenics and support for American individualism.

    Science.gov (United States)

    Cooke, Kathy J

    2002-01-01

    This paper assesses ideas about moral and reproductive duty in American eugenics during the early twentieth century. While extreme eugenicists, including Charles Davenport and Paul Popenoe, argued that social leaders and biologists must work to prevent individuals who were "unfit" from reproducing, moderates, especially Edwin G. Conklin, presented a different view. Although he was sympathetic to eugenic goals and participated in eugenic organizations throughout his life, Conklin realized that eugenic ideas rarely could meet strict hereditary measures. Relying on his experience as an embryologist, Conklin instead attempted to balance more extreme eugenic claims - that emphasized the absolute limits posed by heredity - with his own view of "the possibilities of development." Through his critique he argued that most human beings never even begin to approach their hereditary potential; he moderated his own eugenic rhetoric so that it preserved individual opportunity and responsibility, or what has often been labeled the American Dream.

  11. APA/Psi Chi Edwin B. Newman Graduate Research Award: Meghan H. Puglia.

    Science.gov (United States)

    2016-11-01

    The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and APA. The award was established to recognize young researchers at the beginning of their professional lives and to commemorate both the 50th anniversary of Psi Chi and the 100th anniversary of psychology as a science (dating from the founding of Wundt's laboratory). The 2016 recipient is Meghan H. Puglia, who was chosen for "an outstanding foundational research paper that establishes a relationship between a functional epigenetic modification to the oxytocin receptor gene (OXTR) and neural response during social perception." Puglia's award citation, biography, and bibliography are presented here. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

  12. Evaluation of the integrity of reactor vessels designed to ASME Code, Sections I and/or VIII

    International Nuclear Information System (INIS)

    Hoge, K.G.

    1976-01-01

    A documented review of nuclear reactor pressure vessels designed to ASME Code, Sections I and/or VIII is made. The review is primarily concerned with the design specifications and quality assurance programs utilized for the reactor vessel construction and the status of power plant material surveillance programs, pressure-temperature operating limits, and inservice inspection programs. The following ten reactor vessels for light-water power reactors are covered in the report: Indian Point Unit No. 1, Dresden Unit No. 1, Yankee Rowe, Humboldt Bay Unit No. 3, Big Rock Point, San Onofre Unit No. 1, Connecticut Yankee, Oyster Creek, Nine Mile Point Unit No. 1, and La Crosse

  13. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  14. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  15. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  16. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blotcky, A J; Arsenault, L J [General Medical Research, Veterans Administration Hospital, Omaha (United States)

    1974-07-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  17. Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

    International Nuclear Information System (INIS)

    Wood, RT

    2004-01-01

    This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

  18. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenault, L.J.

    1974-01-01

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  19. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  20. Commercial tandem mirror reactor design with thermal barriers: WITAMIR-I

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Emmert, G.A.; Maynard, C.W.

    1980-10-01

    A conceptual design of a near term commercial tandem mirror power reactor is presented. The basic configuration utilizes yin-yang minimum-B plugs with inboard thermal barriers. The maximum magnetic fields are 6.1 T, 8.1 T, and 15 T in the central cell, yin-yang, and thermal barrier magnets, respectively. The blanket utilizes Pb 83 Li 17 as the coolant and HT-9 as the structural material. This yields a high energy multiplication (1.37), a sufficient tritium breeding ratio (1.07) and has a major advantage with respect to maintenance. The plasma Q is 28 at a fusion power level of 3000 MW(t); the net electrical output is 1530 MW(e); and the overall efficiency is 39%. Cost estimates indicate that WITAMIR-I is competitive with recent tokamak power reactor designs

  1. TRIGASIM: A computer program to simulate a TRIGA Mark I Reactor

    International Nuclear Information System (INIS)

    Ruby, Lawrence

    1992-01-01

    A Fortran-77 computer program has been written which simulates the operation of a TRIGA Mark I Reactor. The 'operator' has options at 1-second intervals, of raising rods, lowering rods, maintaining rods steady, dropping a rod, or scramming the reactor. Results are printed to the screen, and to 2 output files - a tabular record and a logarithmic plot of the power. The Point Kinetic Equations are programmed with 6 delayed groups, quasi-static power feedback, and forward differencing. A pulsing option is available, with simulation which employs the Fuchs Model. A pulse-tail model has been devised to simulate behavior for a few minutes following a pulse. Both graphic and tabular output are also available for the pulses. (author)

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

    International Nuclear Information System (INIS)

    Finn, P.A.

    1985-01-01

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

  3. The physico-chemical 131I species in the exhaust air of a boiling water reactor (BWR 4)

    International Nuclear Information System (INIS)

    Deuber, H.

    1982-12-01

    In a German boiling water reactor, the physico-chemical 131 I species were determined in the plant exhaust and in the individual exhausts during 12 months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental 131 I released to the environment. The retention of the 131 I species by iodine filters was also investigated. On an average, 45% of the 131 I discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhaust from the reactor building and from the turbine building. The other 55% consisted almost entirely of organic I. (orig./HP) [de

  4. Evaluation of I and C architecture alternatives required for the jupiter Icy moons orbiter (JIMO) reactor

    International Nuclear Information System (INIS)

    Muhlheim, M. D.; Wood, R. T.; Bryan, W. L.; Wilson Jr, T. L.; Holcomb, D. E.; Korsah, K.; Jagadish, U.

    2006-01-01

    This paper discusses alternative architectural considerations for instrumentation and control (I and C) systems in high-reliability applications to support remote, autonomous, inaccessible nuclear reactors, such as a space nuclear power plant (SNPP) for mission electrical power and space exploration propulsion. This work supported the pre-conceptual design of the reactor control system for the Jupiter Icy Moons Orbiter (JIMO) mission. Long-term continuous operation without intermediate maintenance cycles forces consideration of alternatives to commonly used active, N-multiple redundancy techniques for high-availability systems. Long space missions, where mission duration can exceed the 50% reliability limit of constituent components, can make active, N-multiple redundant systems less reliable than simplex systems. To extend a control system lifetime beyond the 50% reliability limits requires incorporation of passive redundancy of functions. Time-dependent availability requirements must be factored into the use of combinations of active and passive redundancy techniques for different mission phases. Over the course of a 12 to 20-year mission, reactor control, power conversion, and thermal management system components may fail, and the I and C system must react and adjust to accommodate these failures and protect non-failed components to continue the mission. This requires architectural considerations to accommodate partial system failures and to adapt to multiple control schemes according to the state of non-failed components without going through a complete shutdown and restart cycle. Relevant SNPP I and C architecture examples provide insights into real-time fault tolerance and long-term reliability and availability beyond time periods normally associated with terrestrial power reactor I and C systems operating cycles. I and C architectures from aerospace systems provide examples of highly reliable and available control systems associated with short- and long

  5. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br, E-mail: luquetti@ien.gov.br, E-mail: acerda@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear

    2015-07-01

    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  6. Operation, safety and utilization of the RA reactor in 1978 - Report; Prilog I - Rad, sigurnost i iskoriscenost reaktora RA u 1978. godini - Izvestaj

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-12-15

    This report includes a review of work related to development of reactor operation capacities and increase of the RA reactor safety and economic operation. Statistical data about reactor operation and utilization are included as well. Introducing of the new 80% enriched fuel into the the reactor core enabled increase of the neutron flux, i.e. increase of its production capabilities. Safety and optimization analyses concerned with introduction of the new fuel have shown that the most safe and economic procedure was gradual introducing of the highly enriched fuel. This procedure was based on the concept of mixed core configuration with 2% and 80% enriched fuel elements. By applying this original concept the following significant savings were achieved: fuel elements savings, shortening of the annual period of reactor operation, savings in spent fuel casks, electric power savings, slowing down of heavy water degradation. [Serbo-Croat] Ovaj izvestaj sadrzi pregled o radu na razvoju eksploatacionih mogucnosti i povecanju sigurnosti i ekonomicnosti reaktora RA. Prilozeni su i statisticki podaci o radu i iskoriscenosti reaktora u 1978. godini. Uvodjenje novog 80% obogacenog goriva u jezgro reaktora omogucilo je povecanje neutronskog fluksa tj. povecanje njegovoh proizvodnih mogucnosti. Sigurnosne i optimizacione analize uvodjenja novog goriva pokazale su da je nasigurniji i najekonomicniji postupak postupnog uvodjenja visokoobogacenog goriva koji se zasniva na konceptu mesane resetke sa 2% i 80% obogacenim gorivom. Ovaj originalni koncept omogucio je da se postignu znatne ustede u gorivu, skracivanje godisnjeg rada reaktora, usteda sudova za odlaganje isluzenog goriva, usteda elektricne energije, usporavanje degradacije teske vode.

  7. Study of reactor parameters on the critical systems. Phase I; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N et al [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1962-08-15

    Phase 1 of the report on reactor parameters study describes the preparation of the RB reactor for operation including the following tasks: Completing and verification of reactor safety system; arranging dosimetry instruments; formation of fuel elements with 2% enriched fuel and aluminium holders; improvement of the heavy water level-meter; mounting the horizontal experimental channel; formation of reactor lattice with 16 cm pitch; testing the reactor system; filling the tank with heavy water and preparing the safety report.

  8. RA Reactor operation and maintenance (I-IX), part VII, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    This volume covers the following reports concerned with the maintenance and repair work of the RA reactor: repair of the technical water system; maintenance of the transportation equipment; vacuuming and drying during refurbishment; repair and decontamination of the distillation device; and the report on participation of the operational dosimetry division in the RA reactor refurbishment activities

  9. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  10. Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

  11. I and C safety research at the OECD Halden reactor project

    International Nuclear Information System (INIS)

    Gran, B.A.

    2007-01-01

    The overall objective of the Halden Reactor Project research on software systems dependability is to contribute to the successful introduction of digital I and C systems into NPPs. When celebrating the 50 years of the Halden Project in 2008, about 100 written reports have been delivered within this research. This research covers a number of topics covering safety, reliability, validation and verification, quality assurance, risk assessment, requirement engineering, error propagation, qualitative and quantitative assessment. In the paper some activities are described, pinpointing the importance of good joint projects with organisations in the member countries

  12. The physico-chemical I-131 species in the exhaust air of a boiling water reactor (BWR 5)

    International Nuclear Information System (INIS)

    Deuber, H.

    1984-02-01

    In a German boiling water reactor, the pysico-chemical I-131 species were determined in the plant exhaust and in the individual exhausts during four months. These measurements aimed in particular at determining the percentage and the source of the radiologically decisive elemental I-131 released to the environment. On an average 13% of the I-131 discharged with the plant exhaust consisted of elemental iodine. This was largely released with the exhausts from the reactor building and from the turbine building. The main component was organic-bound I. (orig./HP) [de

  13. Regulations and instructions for RB reactor operation; Propisi i uputstva za rad reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-07-01

    This document includes regulations for reactor RB operation, behaviour and presence of staff in the reactor building; regulations for performing experiments at the RB reactor, regulations and int ructions for the reactor operators and other staff on duty. A chapter is devoted to instruction for reactor operation with the operating documentation and special duties of the operators. Regulations and instruction concerned with accidents are described with classification of accidents and evacuation plan. Annexes to this document include: the present status of the reactor; program for training the reactor operators; forms which are obligatory to be signed for any operating activity, and the certificate of the RB reactor lattice.

  14. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  15. A qualified safety I and C for application in reactors of all kinds

    International Nuclear Information System (INIS)

    Stimler, M.

    2001-01-01

    Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Specific requirements coming from the nuclear qualification have to be fulfilled. To meet both subjects adequately, Siemens has developed the advanced digital I and C technology for safety applications TELEPERM XS. National and international codes and standards impose special requirements on the safety I and C of a nuclear power plant. These concern: fault tolerance; robustness; qualification. In order to be able to meet these requirements to the full without making operational automation tasks unnecessarily expensive by excessive conservatism, the TELEPERM XS I and C system platform was developed. It is largely based on standard Hardware devices selected for their quality characteristics and adapted by specific design measures. In the Software area a complete new development had to be undertaken in order to meet the stringent qualification requirements. In 1992 the GRS (Gesellschaft fuer Reaktorsicherheit - Association for Reactor Safety) confirmed the suitability and licensibility of the underlying TELEPERM XS concepts. Subsequently, the development and qualification of the system software and the engineering tools as well as the type testing of the hardware components was performed. Operationally proven hardware components were selected for utilization, among others from the system families SIMATIC and SINEC. The first integration tests were performed successfully in mid-1996. Field testing of the first application projects could be finalised in 1997. In many countries, the nuclear industry bases its licensing process for nuclear power plants on the US-NRC procedures. For this reason, and in order to ensure world-wide utilization of the TXS technology, it was decided in 1998 to submit a licensing application to the US-NRC. In May 2000, Siemens has received a Safety Evaluation Report (SER) from the US-NRC approving use of its TELEPERM XS (TXS) platform

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  18. Analysis on Configuration of I and C Systems for an Advanced HANARO Reactors

    International Nuclear Information System (INIS)

    Park, Gee Yong; Jung, H. S.; Ryu, J. S.; Park, C.

    2006-01-01

    In an advanced HANARO reactor (AHR), the instrumentation and control (I and C) systems are designed based on the digital system rather than the analog system installed in an existing HANARO instrumentation and control systems. While the safety and functionality of analog-based instrumentation and control system are experienced over a long period of operating time and also well-validated, the obsolescence and the lack of flexibility of this system have to move from the analog technology to the digital technology in the instrumentation and control systems to be used in nuclear power plants as well as nuclear research reactors. For establishing the adequate structure of instrumentation and control systems for an AHR, various instrumentation and control architectures are analyzed for their merits and demerits for use in I and C systems of an AHR and the most promising instrumentation and control architecture for an AHR are drawn from this analysis. The conceptual configuration of a digital-based safety shutdown system is proposed in this report

  19. Feasibility study for production of I-131 radioisotope using MNSR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Elom Achoribo, A.S., E-mail: achoribo@yahoo.fr [Radiological and Medical Sciences Research Institute, Ghana Atomic Energy Commission, P.O. Box LG80, Legon Accra (Ghana); Akaho, Edward H.K. [Ghana Atomic Energy Commission, P.O. Box LG80, Legon Accra (Ghana); Nyarko, Benjamin J.B.; Osae Shiloh, K.D. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG80, Legon Accra (Ghana); Odame Duodu, Godfred [Radiological and Medical Sciences Research Institute, Ghana Atomic Energy Commission, P.O. Box LG80, Legon Accra (Ghana); Gibrilla, Abass [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG80, Legon Accra (Ghana)

    2012-01-15

    A feasibility study for {sup 131}I production using a Low Power Research Reactor was conducted to predict the yield of {sup 131}I by cyclic activation technique. A maximum activity of 5.1 GBq was achieved through simulation using FORTRAN 90, for an irradiation of 6 h. But experimentally only 4 h irradiation could be done, which resulted in an activity of 4.0 Multiplication-Sign 10{sup 5} Bq. The discrepancy in the activities was due to the fact that beta decays released during the process could not be considered. - Highlights: Black-Right-Pointing-Pointer For a high irradiation time, the neutron flux will give high activity. Black-Right-Pointing-Pointer For maximum number of irradiation that can be done a maximum activity could be obtained. Black-Right-Pointing-Pointer An idea on how to maximize the activity (recommendation).

  20. Investigation of efficient {sup 131}I production from natural uranium at Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khalafi, H. [Nuclear Research Center, AEOI, No. 54 North Kargar Avenue, P.O. Box 14155/1339, Tehran (Iran, Islamic Republic of)]. E-mail: hossein_khalafi@yahoo.com; Nazari, K. [Jaber-Ibne-Hayan Research laboratories, AEOI, P.O. Box 11365/8486, Tehran (Iran, Islamic Republic of); Ghannadi-Maragheh, M. [Jaber-Ibne-Hayan Research laboratories, AEOI, P.O. Box 11365/8486, Tehran (Iran, Islamic Republic of)

    2005-05-15

    Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of {sup 131}I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO{sub 2}) for 100 h at 3.5 x 10{sup 13} (n's/cm{sup 2} s) thermal neutron flux in the TRR, one can produce about 5 Ci of {sup 131}I for medical purposes, on the other hand can produce very useful radionuclides like {sup 99}Mo and {sup 133}Xe in one batch irradiation in the unique production line.

  1. Investigation of efficient 131I production from natural uranium at Tehran research reactor

    International Nuclear Information System (INIS)

    Khalafi, H.; Nazari, K.; Ghannadi-Maragheh, M.

    2005-01-01

    Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of 131 I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO 2 ) for 100 h at 3.5 x 10 13 (n's/cm 2 s) thermal neutron flux in the TRR, one can produce about 5 Ci of 131 I for medical purposes, on the other hand can produce very useful radionuclides like 99 Mo and 133 Xe in one batch irradiation in the unique production line

  2. Using level-I PRA for enhanced safety of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Ramsey, C.T.; Linn, M.A.

    1995-01-01

    The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

  3. RB Research nuclear reactor, Annual report for 1995, I-IV

    International Nuclear Information System (INIS)

    Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

    1995-12-01

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor

  4. RB Research nuclear reactor, Annual report for 1996, I-IV

    International Nuclear Information System (INIS)

    Stefanovic, D.; Milosevic, M.; Pesic, M.; Marinkovic, P.; Ilic, R.; Dasic, N.; Milovanovic, S.; Ljubenov, V.; Petronijevic, M.; Jevremovic, M.

    1996-12-01

    Report on RB reactor operation during 1996 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a list of publications resulting from experiments done at the RB reactor

  5. The physico-chemical 131I species in the stack exhaust air of a boiling water reactor

    International Nuclear Information System (INIS)

    Deuber, H.

    1982-07-01

    In the stack exhaust air of a German boiling water reactor, the fractions of elemental, particulate and organic 131 I were determined over a period of three years. The average fraction of elemental 131 I, which is decisive for the ingestion dose, was about 20% during the first two years and about 50% during the third year. (orig.) [de

  6. RA reactor operation and maintenance in 1994, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1994. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1994-12-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Zapoceti su radovi na modernizaciji, odnosno zameni instrumentacije reaktora ali njegova realizacija kasni

  7. Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.

    1976-12-01

    In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

  8. Edwin Austin Abbey's The Passage of the Hours: Astronomy as History

    Science.gov (United States)

    Ricci, P. L.

    2016-01-01

    The Passage of the Hours (1909-1911) in the Pennsylvania State Capitol at Harrisburg is one of the most original and least known astronomical ceilings in the United States. Designed by the American artist Edwin Austin Abbey (1852-1911) to complement the Italian Renaissance style architecture of the House of Representatives, the mural combines two classical traditions of representing the night sky: a celestial map with the constellations of the zodiac and the personifications of the Hours. Set in a shallow dome twenty-four feet in diameter, Abbey's constellation figures float in a dazzling firmament where the Milky Way streams between the Sun and the Moon. The artist placed the Horae of Greek mythology around the dome's circumference in the position of the numbers on an astronomical clock. In the tradition of Italian Renaissance architecture, the celestial ceiling in the House of Representatives was part of an iconographic program affirming the cosmological origin of a polity. The astronomical theme relates to Abbey's murals in the House Chamber of the first public reading of the Declaration of Independence in 1776 from David Rittenhouse's observatory in Philadelphia, which the astronomer constructed to study the transit of Venus in 1769. The artist included a portrait of Rittenhouse holding his telescope among the worthies in the adjacent mural of The Apotheosis of Pennsylvania. Contemporary as well as historical events encouraged Abbey's use of astronomical imagery: the depiction of a comet may record the much-anticipated return of Halley's Comet in 1910.

  9. Sümeyra Tosun: Psi Chi/APA Edwin B. Newman Graduate Research Award.

    Science.gov (United States)

    2014-11-01

    The Edwin B. Newman Graduate Research Award is given jointly by Psi Chi and APA. The award was established to recognize young researchers at the beginning of their professional lives and to commemorate both the 50th anniversary of Psi Chi and the 100th anniversary of psychology as a science (dating from the founding of Wundt's laboratory). The 2014 recipient is Sümeyra Tosun. Tosun was chosen for "an outstanding research paper that examines the cognitive repercussions of obligatory versus optional marking of evidentiality, the linguistic coding of the source of information. In English, evidentiality is conveyed in the lexicon through the use of adverbs. In Turkish, evidentiality is coded in the grammar. In two experiments, it was found that English speakers were equally good at remembering and monitoring the source of firsthand information and the source of non-firsthand information. Turkish speakers were worse at remembering and monitoring non-firsthand information than firsthand information and were worse than English speakers at remembering and monitoring non-firsthand information." Tosun's award citation, biography, and a selected bibliography are presented here. PsycINFO Database Record (c) 2014 APA, all rights reserved.

  10. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987; Istrazivacki nuklearni reaktor RA Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1987. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1987-12-15

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction. [Serbo-Croat] Reaktor RA nije radio usled zabrane Izvsnog veca Skupstine Srbije od 27. avgusta 1984. U cilju povecanja pouzdanosti rada reaktora a da bi se udovoljilo zakonskim propisima sto je uslov za dobijanje stalne dozvole za rad realizovana su tri velika zahvata na reaktoru RA. Ovi zahvati obuhvatili su izgradnju sistema za hladjenje jezgra reaktora u slucaju nuzde, rekonstukciju postojeceg sistema specijalne ventilacije i rekonstrukciju sistema napajanja elektricnom energijom neophodnih potrosaca reaktora RA. Zapoceti su radovi na modernizaciji intrumentacije reaktora RA, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, a trebalo bi da se realizuje do kraja 1989. godine. U cilju povecanja prostora za skladistenje ozracenog nuklearnog goriva i njegovog efikasnijeg koriscenja, izradjen je su projekti za rekonstrukciju postojecih uredjaja za rukovanje gorivom, povecanje smestajnog kapaciteta i preciscavanje vode u bazenima za odlezavanje. Realizaija ovih

  11. Transmutation of Tc-99 and I-129 in fission reactors. A calculational study

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1995-03-01

    The HWR is a better candidate for large-scale transmutation of long-lived fission products. When target pins containing either Tc-99 or I-129 are positioned in the centre of each fuel bundle of a 935 MW e CANDU reactor, the transmutation half lives are 44 and 20 years, respectively, and the gross transmutation rates 60 and 48 kg/a. The positive coolant void coefficient is reduced in both cases with about 30%. When Tc-99 target pins are positioned in the moderator between the fuel bundles, the transmutation half life becomes 25 years and the gross transmutation rate 106 kg/a. This means that one HWR can serve four PWRs with equal power. The fast reactor seems most promising. When Tc-99 target pins are irradiated in moderated subassemblies in the inner core of Superphenix (∼1240 MW e ), a transmutation half life of 15 years is obtained with a gross transmutation rate of 122 kg/a. These values become 18 years and 101 kg/a when non-moderated subassemblies are used for the irradiation. This implies that one fast reactor can serve four to five PWRs with equal power. The PWR seems not very effective for transmutation of Tc-99. Large inventories are needed to obtain a Tc-99 transmutation rate equal to the production rate (18 kg/a for a 900 MW e PWR). When all guide tubes of an UO 2 fuelled PWR are filled with Tc-99 with density of 5 g cm -3 , the transmutation half life is 39 years and the gross transmutation rate 64 kg/a. (orig./GL)

  12. Maintenance procedures for the TITAN-I and TITAN-II reversed field pinch reactors

    International Nuclear Information System (INIS)

    Grotz, S.P.; Duggan, W.; Krakowski, R.; Najmabadi, F.; Wong, C.P.C.

    1989-01-01

    The TITAN reactor is a compact, high-power-density (neutron wall loading 18 MW/m 2 ) machine, based on the reversed-field-pinch (RFP) confinement concept. Two designs for the fusion power core have been examined: TITAN-I is based on a self-cooled lithium loop with a vanadium-alloy structure for the first wall, blanket and shield; and TITAN-II is based on an aqueous loop-in-pool design with a LiNO 3 solution as the coolant and breeder. The compact design of the TITAN fusion power core, (FPC) reduces the system to a few small and relatively low mass components, making toroidal segmentation of the FPC unnecessary. A single-piece maintenance procedure is possible. The potential advantages of single-piece maintenance procedures are: (1) Short period of down time; (2) improved reliability; (3) no adverse effects resulting from unequal levels of irradiation; and (4) ability to continually modify the FPC design. Increased availability can be expected from a fully pre-tested, single-piece FPC. Pre-testing of the FPC throughout the assembly process and prior to installation into the reactor vault is discussed. (orig.)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  15. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora RA (I-IX), IV Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems.

  16. 131I content in canine thyroids in the Warsaw urban area after the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Krauze, S.; Rozycki, Z.; Sitarska, E.

    1987-01-01

    The levels of 131 I were determined in the thyroids of 20 dogs from Warsaw submitted to euthanasia between May and September 1986. The animals were living with humans and were in similar way exposed to contamination after the Chernobyl reactor accident. After calculation of the radioactivity for May 10th the contamination was found to range from 142.9 to 1372.9 Bq. These values corresponded to the contamination of human thyroids as reported by Central Laboratory for Radiation Protection in Warsaw. From the begining of May to the end of November the number of operations performed in dogs for pathological thyroid hyperplasia was six times higher than in the preceding time period. 5 refs., 2 tabs. (author)

  17. Preliminary evaluation of the stress analysis reports for Angra I reactor coolant loop - part 1

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.; Andrade, J.E.L. de

    1980-03-01

    A methodology that will allow CNEN to approve the stress analysis reports of the components of the Brazilian nuclear power plants, was developed. The reactor coolant loop (RCL)of Angra I was checkd. This is the first part of the complete report and consists of the approval of the design documents, the approval of the equipment support models and the aproval of the steam generator dynamic model. The second part of this work is under way now and should contain the approval of the RCL stress and fatigue analysis according to ASME code section III. As shown in section 7 it appears necessary additional information from Westinghouse about the design of the RCL. (Author) [pt

  18. Requalification of the LOFT reactor following a loss of coolant experiment (Level I)

    International Nuclear Information System (INIS)

    Cannon, J.W.

    1979-01-01

    During a Loss of Coolant Experiment (LOCE), the LOFT reactor experiences an acceleration of 10 G's and fuel cladding temperature changes at a rate of 1100 0 K/sec. These unparalleled conditions present a unique startup problem to the LOFT program: How can the integrity of the fuel be confirmed so as to minimize operation if damage has occurred. The Level I Requalification Program is designed to accomplish this. It is a progressive series of tests, designed to detect damage at the earliest possible time, and thus preclude or minimize operation if damage exists. First, fuel specialists examine the LOCE data for possible damaging conditions and the results of primary coolant sample analysis for signs of failed fuel. Second, the requalification program proceeds to a series of mechanical and physics tests

  19. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  20. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  1. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    Ventura, Mirta A.; Rosso, Ricardo D.

    2002-01-01

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  2. Study of advanced fission power reactor development for the United States. Volume I

    International Nuclear Information System (INIS)

    1976-01-01

    This volume summarizes the results and conclusions of an assessment of five advanced fission power reactor concepts in the context of potential nuclear power economies developed over the time period 1975 to 2020. The study was based on the premise that the LMFBR program has been determined to be the highest priority fission reactor program and it will proceed essentially as planned. Accepting this fact, the overall objective of the study was to provide evaluations of advanced fission reactor systems for input to evaluating the levels of research and development funding for fission power. Evaluation of the reactor systems included the following categories: (1) power plant performance, (2) fuel resource utilization; (3) fuel-cycle requirements; (4) economics; (5) environmental impact; (6) risk to the public; and (7) R and D requirements to achieve commercial status. The specific major objectives of the study were twofold: (1) to parametrically assess the impact of various reactor types for various levels of power demand through the year 2020 on fissile fuel utilization, economics, and the environment, based on varying but reasonable assumptions on the rates of installation; and (2) to qualitatively assess the practicality of the advanced reactor concepts, and their research and development. The reactor concepts examined were limited to the following: advanced high-temperature, gas-cooled reactor (HTGR) systems including the thorium/U-233 fuel cycle, gas turbine, and binary cycle (BIHTGR); gas-cooled fast breeder reactor (GCFR); molten salt breeder reactor (MSBR); light water breeder reactor (LWBR); and CANDU heavy water reactor

  3. RA Reactor operation and maintenance (I-IX), part VIII, Task 3.08/05, Decontamination of the reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    Permanent increase of radiation in the heavy water system was noticed during first three year of the RA reactor operation, even when the reactor was shutdown. It was found that there was no failure of the fuel element cladding. Radioactive cobalt was found in the heavy water which was rather strange. During repair of the heavy water system, it has been found that stellite was used for coating the heavy water pumps. Since stellite is a cobalt alloy, this could have been the source of radioactive cobalt in the heavy water. The stellite coating was damaged due to friction and particle of cobalt appeared in the coolant, they were activated since they were in the core. decontamination of the heavy water and the heavy water coolant loop was a must . Beside the detailed report on the contamination and decontamination of the heavy water system this volume includes 14 annexes describing the investigation of the event and the whole procedure of decontamination

  4. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  5. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

  6. Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S n code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S 4 approximation. Reactor cell was treated as D 2 O-U-D 2 O system. Time of iteration was 185 s [sr

  7. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-10-15

    At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

  8. RA reactor operation in 1991, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1991. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1991, three major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. Renewal of the reactor instrumentation was started but but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA for political reasons. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Training of the existing personnel was done regularly, but lack of financial support prevented employment of new workers that would be needed for operation in shifts and regular maintenance. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Kako bi se ubuduce mogao obezbediti

  9. A New Deal for Southeastern Archaeology, by Edwin A. Lyon, The University of Alabama Press

    Directory of Open Access Journals (Sweden)

    Gordon R. Willey

    1996-05-01

    Full Text Available In the 1950s, the Era of the Great Depression, archaeology in the United States enjoyed an enormous boost, both in the substance of its findings on the Precolumbian past and in the development of its methods and proce­dures. Edwin A. Lyon has laid out the story of all this in a book that is a major contribution to the history of the archaeological discipline in this country. The context of this story is in the American South, most specifically the Southeastern United States, or the 'Old South', that part of the country that was the heart of the Confederacy; and it is important to remember that the South has had a history significantly separate and distinct from that of the rest of the nation. This separateness, rooted in its plantation economy and the associated institution of slavery, was further fostered by the Civil War and its aftermath of hardships. These hardships lasted until the 1930s and the economic depression when they began to be ameliorated by the Rooseveltian political and socioeconomic measures known collectively as the 'New Deal'. The policies of the New Deal began those transformations which continued through World War II and beyond. Crucial to these transformations were the building of power dams and rural electrification, soil erosion control and agricultural modernization, and a host of public building programs. All of this went forward with Federal Relief employment. Less tangible but nonetheless important benefits were in the cultural sphere: the arts, drama, writing. history - and of particular importance to us here. archaeology.

  10. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  11. Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

    International Nuclear Information System (INIS)

    Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

    2010-01-01

    These days, the demand to use nuclear resources efficiently is growing for long-term energy supply and also for solving the green house problem. It is indispensable to develop technologies to reduce environmental load with the nuclear energy supply for sustainable development of human beings. In this regard, the development of the fast breeder reactor (FBR) is preferable to utilize nuclear resources effectively and also to burn minor actinides which possess very long toxicity for more than thousands years if they are not extinguished. As one of the FBR developing works in Japan this phase I study started in 2006 to introduce hafnium (Hf) hydride and Gadolinium-Zirconium (Gd-Zr) hydride as new control materials in FBR. By adopting them, the FBR core control technology is improved by two ways. One is extension of control rod life time by using long life Hf hydride which leads to reduce the fabrication and disposal cost and the other is reduction of the excess reactivity by adopting Gd-Zr hydride which leads to reduce the number of control rods and simplifies the core upper structure. This three year study was successfully completed and the following results were obtained. The core design was performed to examine the applicability of the Hf hydride absorber to Japanese Sodium Fast Reactor (JSFR) and it is clarified that the control rod life time can be prolonged to 6 years by adopting Hf hydride and the excess reactivity of the beginning of the core cycle can be reduced to half and the number of the control rods is also reduced to half by using the Gd-Zr hydride burnable poison. The safety analyses also certified that the core safety can be maintained with the same reliability of JSFR Hf hydride and Gd-Zr hydride pellets were fabricated in good manner and their basic features for design use were measured by using the latest devices such as SEM-EDX. In order to reduce the hydrogen transfer through the stainless steel cladding a new technique which shares calorizing

  12. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  13. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Manimaran, M.; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-01-01

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored

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

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto

    2008-01-01

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

  15. First experiences from system integration, installation and commissioning of TELEPERM XS for reactor I and C at the Unterweser NPP

    International Nuclear Information System (INIS)

    Schoerner, O.

    1998-01-01

    The modernization of Reactor I and C, consisting of reactor limitation system, reactor control system and rod control system, at Unterweser NPP is the pilot application of the state-of-the-art safety I and C system TELEPERM XS. The Unterweser system has been integrated and tested from December 1996 to May 1997 in the Siemens Erlangen test field and has been installed at site in July 1997. For the period from July 1997 to Jul 1998 the new TELEPERM XS based Reactor I and C system will be operated online-open-loop in parallel to the existing system, in order to get information about the long term stability of the system and conduct intensive personnel training. For one selected function ''Power distribution control'' the operator has the possibility to choose between the old controller and the new TELEPERM XS function. During the 1998 outage the TELEPERM XS system will be connected to the process and the old I and C system will be dismantled. This document describes the experiences gathered during system integration in the test field. (author)

  16. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1983. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Raickovic, N; Radivojevic, J; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1983-12-15

    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. [Serbo-Croat] Pregledom nuklearnog goriva iz tehnoloskih kanala reaktora RA koji je izvrsen u periodu decembear 1982-feburuar 1983. godine nakon zaustavljanja reaktora po isteku novembarske kampanje 1982. godine, ustanovljeno je da ponovo dolazi do stvaranja taloga u obliku hidratisanih oksida aluminiuma na kosuljicama gorivnih elemenata. Nakon ponovnog pustanja u rad, reaktor je do novembra 1981. godine neprekidno bio u pogonu na snagama 1,8 - 2 MW. Jezgro je bilo formirano iskljucivo sa od gorivnih elemenata sa 80% obogacenim uran dioksidom. Utvrdjeno je da kosuljica gorivnog elementa nije korodirala, i da se bez nastval rada ne moze utvrditi uzrok pojave taloga na povrsini kosuljice. Da bi se mogucnost korozije aluminjumskih komponenti u primarnom kolu raktora svela na sto manju meru odluceno je da se vrsi neprekidno preciscavanje teske vode i da se istovremeno poveca protok teske vode radom dve pumpe, Komitet za sigurnost Instituta odobrio je ovakav nacin

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-12-15

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

  18. Fault tolerant distributed real time computer systems for I and C of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2014-03-15

    Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network.

  19. Operation and maintenance of RA Reactor, Annual report 1977; Pogon i odrzavanje reaktora RA - Izvestaj o radu u 1977. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    During 1977, the RA Reactor was operated at nominal power of 6.5 MW for 183 days. Total production was 28582 MWh which is 10% higher than planned. Second phase of introducing the 80% enriched fuel was fulfilled according to the plan. This means that the reactor core will be filled with highly enriched fuel in 1978. Refueling was done three time during the past year. After completing the first phase of the fuel exchange which was related mostly to reactor safety, the second phase will be devoted to the more efficient increase of neutron flux . This second phase is of utmost importance because higher neutron flux will provide better and more efficient reactor application from economic point of view. This will justify the application of the new more expensive highly enriched fuel. The budget for reactor operation and maintenance is hardly enough to cover the maintenance of the components and instrumentation. During 1977 there were no accidents related nor incidents related to the instrumentation or related to radiation protection. [Serbo-Croat] Reaktor RA je u toku 1977. godine ostvario rad od 28583 MWh odnosno 183 dana rada na nominalnoj snazi, sto u odnosu na plan rada iznosi 10% vise od planiranog. Druga faza uvodjenja 80% obogacenog goriva izvrsena je prema planu sto znaci da ce se punjenje jezgra reaktora 80% gorivom zavrsiti u toku 1978. godine. Izvrsene su tri izmene goriva. Posle isteka prvog dela prelaznog rezima koji je usmeren na maksimalnu sigurnost reaktora preci ce se na drugi deo usmeren na vece i brze povecanje neutronskog fluksa. Druga faza je od izuzetnog znacaja jer ce omoguciti bolje i znatno ekonomicnije koriscenje reaktora i opravdati upotrebu novog i skupljeg goriva. Sredstva za pogon i odrzavanje reaktora RA su jedva dovoljna za odrzavanje neophodnog nivoa opreme. Akcidenata u toku 1977. godine nije bilo ni sa opremom ni u pogledu zastite od zracenja.

  20. Advanced I&C for Fault-Tolerant Supervisory Control of Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Daniel G. [Univ. of Pittsburgh, PA (United States)

    2018-01-30

    In this research, we have developed a supervisory control approach to enable automated control of SMRs. By design the supervisory control system has an hierarchical, interconnected, adaptive control architecture. A considerable advantage to this architecture is that it allows subsystems to communicate at different/finer granularity, facilitates monitoring of process at the modular and plant levels, and enables supervisory control. We have investigated the deployment of automation, monitoring, and data collection technologies to enable operation of multiple SMRs. Each unit's controller collects and transfers information from local loops and optimize that unit’s parameters. Information is passed from the each SMR unit controller to the supervisory controller, which supervises the actions of SMR units and manage plant processes. The information processed at the supervisory level will provide operators the necessary information needed for reactor, unit, and plant operation. In conjunction with the supervisory effort, we have investigated techniques for fault-tolerant networks, over which information is transmitted between local loops and the supervisory controller to maintain a safe level of operational normalcy in the presence of anomalies. The fault-tolerance of the supervisory control architecture, the network that supports it, and the impact of fault-tolerance on multi-unit SMR plant control has been a second focus of this research. To this end, we have investigated the deployment of advanced automation, monitoring, and data collection and communications technologies to enable operation of multiple SMRs. We have created a fault-tolerant multi-unit SMR supervisory controller that collects and transfers information from local loops, supervise their actions, and adaptively optimize the controller parameters. The goal of this research has been to develop the methodologies and procedures for fault-tolerant supervisory control of small modular reactors. To achieve

  1. Dense Medium Plasma Water Purification Reactor (DMP WaPR), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

  2. Analysis of dynamic stability and safety of reactor system by reactor simulator; Analiza dinamicke stabilnosti i sigurnosti reaktorskog sistema pomocu reaktorskog simulatora

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-11-15

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor.

  3. A strategy to apply a graded approach to a new research reactor I and C design

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Park, Jae Kwan; Kim, Taek Kyu; Bae, Sang Hoon; Baang, Dane; Kim, Young Ki

    2012-01-01

    A project for the development of a new research reactor (NRR) was launched by KAERI in 2012. It has two purposes: 1) providing a facility for radioisotope production, neutron transmutation doping, and semiconductor wafer doping, and 2) obtaining a standard model for exporting a research reactor (RR). The instrumentation and control (I and C) design should reveal an appropriate architecture for the NRR export. The adoption of a graded approach (GA) was taken into account to design the I and C and architecture. Although the GA for RRs is currently under development by the IAEA, it has been recommended and applied in many areas of nuclear facilities. The Canadian Nuclear Safety Commission allows for the use of a GA for RRs to meet the safety requirements. Germany applied the GA to a decommissioning project. It categorized the level of complexity of the decommissioning project using the GA. In the case of 10 C.F.R. Part 830 830.7, a contractor must use a GA to implement the requirements of the part, document the basis of the GA used, and submit that document to U.S. DOE. It mentions that a challenge is the inconsistent application of GA on DOE programs. RG 1.176 states that graded quality assurance brings benefits of resource allocation based on the safety significance of the items. The U.S. NRC also applied the GA to decommissioning small facilities. The NASA published a handbook for risk informed decision making that is conducted using a GA. ISATR67.04.09 2005 supplements ANSI/ISA.S67.04.01. 2000 and ISA RP67.04.02 2000 in determining the setpoint using a GA. The GA is defined as a risk informed approach that, without compromising safety, allows safety requirements to be implemented in such a way that the level of design, analysis, and documentation are commensurate with the potential risks of the reactor. The IAEA is developing a GA through DS351 and has recommended applying it to a reactor design according to power and hazarding level. Owing to the wide range of RR

  4. A strategy to apply a graded approach to a new research reactor I and C design

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Suk; Park, Jae Kwan; Kim, Taek Kyu; Bae, Sang Hoon; Baang, Dane; Kim, Young Ki [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    A project for the development of a new research reactor (NRR) was launched by KAERI in 2012. It has two purposes: 1) providing a facility for radioisotope production, neutron transmutation doping, and semiconductor wafer doping, and 2) obtaining a standard model for exporting a research reactor (RR). The instrumentation and control (I and C) design should reveal an appropriate architecture for the NRR export. The adoption of a graded approach (GA) was taken into account to design the I and C and architecture. Although the GA for RRs is currently under development by the IAEA, it has been recommended and applied in many areas of nuclear facilities. The Canadian Nuclear Safety Commission allows for the use of a GA for RRs to meet the safety requirements. Germany applied the GA to a decommissioning project. It categorized the level of complexity of the decommissioning project using the GA. In the case of 10 C.F.R. Part 830 830.7, a contractor must use a GA to implement the requirements of the part, document the basis of the GA used, and submit that document to U.S. DOE. It mentions that a challenge is the inconsistent application of GA on DOE programs. RG 1.176 states that graded quality assurance brings benefits of resource allocation based on the safety significance of the items. The U.S. NRC also applied the GA to decommissioning small facilities. The NASA published a handbook for risk informed decision making that is conducted using a GA. ISATR67.04.09 2005 supplements ANSI/ISA.S67.04.01. 2000 and ISA RP67.04.02 2000 in determining the setpoint using a GA. The GA is defined as a risk informed approach that, without compromising safety, allows safety requirements to be implemented in such a way that the level of design, analysis, and documentation are commensurate with the potential risks of the reactor. The IAEA is developing a GA through DS351 and has recommended applying it to a reactor design according to power and hazarding level. Owing to the wide range of RR

  5. RA reactor operation and maintenance in 1989, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1989. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Krajem 1988, medjunarodna agencija za atomsku energiju potpisala je ugovor sa sovjetskom firmom Atomergexport za izradu novog sistema instrumentacije. Sa ciljem da se poveca i efikasnije koristi prostor za skladistenje ozracenog goriva, 1987. godine zapoceta je realizacija projekata preciscavanja vode u bazenima za odlezavanje

  6. Monolitni katalizatori i reaktori: osnovne značajke, priprava i primjena (Monolith catalysts and reactors: preparation and applications

    Directory of Open Access Journals (Sweden)

    Tomašić, V.

    2004-12-01

    Full Text Available Monolithic (honeycomb catalysts are continuous unitary structures containing many narrow, parallel and usually straight channels (or passages. Catalytically active components are dispersed uniformly over the whole porous ceramic monolith structure (so-called incorporated monolithic catalysts or are in a layer of porous material that is deposited on the walls of channels in the monolith's structure (washcoated monolithic catalysts. The material of the main monolithic construction is not limited to ceramics but includes metals, as well. Monolithic catalysts are commonly used in gas phase catalytic processes, such as treatment of automotive exhaust gases, selective catalytic reduction of nitrogen oxides, catalytic removal of volatile organic compounds from industrial processes, etc. Monoliths continue to be the preferred support for environmental applications due to their high geometric surface area, different design options, low pressure drop, high temperature durability, mechanical strength, ease of orientation in a reactor and effectiveness as a support for a catalytic washcoat. As known, monolithic catalysts belong to the class of the structured catalysts and/or reactors (in some cases the distinction between "catalyst" and "reactor" has vanished. Structured catalysts can greatly intensify chemical processes, resulting in smaller, safer, cleaner and more energy efficient technologies. Monolith reactors can be considered as multifunctional reactors, in which chemical conversion is advantageously integrated with another unit operation, such as separation, heat exchange, a secondary reaction, etc. Finally, structured catalysts and/or reactors appear to be one of the most significant and promising developments in the field of heterogeneous catalysis and chemical engineering of the recent years. This paper gives a description of the background and perspectives for application and development of monolithic materials. Different methods and techniques

  7. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976; Pogon i odrzavanje reaktora RA, Izvestaj o radu u 1976. godini

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation. [Serbo-Croat] Reaktor RA je u 1976. godini radio za oko 30% krace vreme od uobicajenog. Razlog su bill izuzetno veliki obim remontnih i drugih radova u okviru tekuceg i investicionog odrzavanja, kao i krupnih

  8. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  9. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  10. Operating history report for the Peach Bottom HTGR. Volume I. Reactor operating history

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Baldwin, N.L.; Tomlin, R.W.

    1976-01-01

    The operating history for the Peach Bottom-1 Reactor is presented for the years 1966 through 1975. Information concerning general chemistry data, general physics data, location of sensing elements in the primary helium circuit, and postirradiation examination and testing of reactor components is presented

  11. Radiation control and safety of fast reactor; Radijaciona kontrola i sigurnost postrojenja sa brzim reaktorom

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Antic, D [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1983-07-01

    The fundamental activities for safeguard of radiation control and safety and the necessary staff for them for fast reactor plant are shown. The basic systems for the plant radiation control are counted, especially with regards to poisoning of some fuel materials. The particular characteristics of the plant radiation control determined by the fast reactor are pointed out. (author)

  12. Decontamination and decommissioning of the SPERT-I Reactor Building at the Idaho National Engineering Laboratory. Final report

    International Nuclear Information System (INIS)

    Dolenc, M.R.

    1986-02-01

    This final report documents the decontamination and decommissioning of the SPERT-I Reactor Building. This 20- by 40-ft galvanized steel building was dismantled; and the resultant contaminated sludge, liquid, and carbon steel were disposed of at the Radioactive Waste Management Complex of the Idaho National Engineering Laboratory. This report presents the results of the characterization, decision analysis, planning, and decommissioning of the facility. The total cost of these activities was $139,500. Of this total, $103,500 was required for decommissioning operations. (This latter figure represents a 20% savings over the estimated costs generated during the planning effort.) The objectives of decommissioning this facility were to stabilize the seepage pit area and remove the reactor building. The D and D work was divided into two parts; the seepage pit was decommissioned in 1984, and the reactor building in 1985. The entire area was backfilled with radiologically clean soil, graded, and seeded. Two markers were installed to identify the locations of the pit and reactor building. The only isotopes found in either decommissioning operation were cesium-137 and uranium-235 in very low concentrations. Decommissioning operations of the reactor building were carried out during August 1985. The project generate 297 ft 3 of radioactive waste. No personnel radiation exposure above background was received by D and D workers

  13. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2009-03-15

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  14. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    International Nuclear Information System (INIS)

    Seifert, H.-P.; Ritter, S.

    2009-03-01

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  15. Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5; Prilog 5 - Dozimetrija i tehnicka zastita

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-15

    In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation. [Serbo-Croat] Na reaktoru RA vrsi se stalna kontrola nivoa gama zracenja po tehnoloskim i radnim prostorijama, pomocu ugradjenog stacionarnog sistema. Svakodnevno se vrse merenja u reaktorskoj hali, prema ukazanim potrebama. Posebno se mere nivoi gamma zracenja u karakteristicnim tackama pri radu reaktora na nominalnoj snazi i u pauzama izmedju dva kampanje. Merenje nivoa neutronskog zracenja vrsi se diskontinualno pomocu mobilnog sfernog neutronskog dozimetra. Merenja se obavljaju u hali oko horizontalnih eksperimentalnih kanala. Izmerene vrednosti su u proseku tri puta manje od odgovarajucih nivoa gama zracenja.

  16. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon [Youngdong Univ., Yeongdong (Korea, Republic of)] (and others)

    2003-03-15

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study.

  17. A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

    International Nuclear Information System (INIS)

    Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon

    2003-03-01

    In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study

  18. Determination of the effective delayed neutron fraction in the Coral-I Reactor

    International Nuclear Information System (INIS)

    Francisco, J. L. de; Perez-Navarro, A.; Rodriguez-Mayquez, E.

    1973-01-01

    The effective delayed neutron fraction, β eff, has been determined from the measurement of E / β 2 , by means of reactor noise analysis in the time domain, and the neutron detector efficiency, ε. For the ε measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm 3 , at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

  19. Spare parts management for nuclear research reactors [Paper No.: I-14

    International Nuclear Information System (INIS)

    Kini, M.P.

    1981-01-01

    Most of the equipment installed at CIRUS and other reactors are imported units. CIRUS reactor is 20 years old and with present problems for obtaining spare parts for this equipment, indigenous effort in procurement has become imperative. In the absence of specifications and drawings for most of the components, the task of indigenous procurement has become quite demanding. The efforts put by Reactor Operations Division of the Bhabha Atomic Research Centre, Bombay in locating local manufacturers who are willing to fabricate in small quantities of spare parts to specifications and the difficulties involved is the theme of this paper. The paper also covers the efforts on equipment replacement, its success and failures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-07-01

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

  1. Design of emergency shutdown system for the Tehran Research Reactor; Part I: Neutronics investigation

    International Nuclear Information System (INIS)

    Safarinia, M.; Faghihi, F.; Mirvakili, S.M.; Fakhraei, A.

    2017-01-01

    Highlights: • An emergency shutdown system for the TRR is carried out based on a heavy water tank. • The performance of the heavy water tank are carried out based on “first and equilibrium cores”. • Heavy water discharging flow rate is also studied in the current research. • Thermal flux in the radioisotope channel with and without the heavy water tank are studied. • A core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). - Abstract: In this paper, a neutronics design of the secondary (i.e., emergency) shutdown system for the Tehran Research Reactor (TRR) is carried out based on a heavy water tank design. The heavy water tank in a cylindrical shape is around the core, and calculations for the optimized radius and height of the tank are performed. The performance of the heavy water tank calculations are carried out based on two types of fuel loading, which are called the “first and equilibrium cores” of the TRR. For both cases, neutronics and standard safety analysis are taken into account, benchmarked, and described herein. Heavy water discharging flow rate is also studied in the current research, and the results are compared with the IAEA criteria. Moreover, thermal flux in the radioisotope channel with and without the heavy water tank (as the reflector) are studied herein. Specifically, a core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). Based on our optimization, the 5 × 5 fuel assembly, which is called “B configuration,” has better performance and efficiency than that of the other described layouts.

  2. Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Lerner, Ana M.; Madariaga, Marcelo R.

    1999-01-01

    In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)

  3. Novel, Regenerable Microlith Catalytic Reactor for CO2 Reduction via Bosch Process, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Precision Combustion, Inc. (PCI) proposes to develop an extremely compact, lightweight and regenerable MicrolithREG catalytic CO2 reduction reactor, capable of...

  4. Operation and maintenance of the RA reactor; Pogon i odrzavanje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    During 1961 the RA reactor was used for irradiation of samples for 258 users ( from which 188 from the Institute) and for 34 experiments, this means productions of 5958 MWh. The production was almost four times bigger than during previous year due to the demand of the Laboratories of the Institute. Burnup of the fuel from the first batch was about 30%. This means that the operation time could be much higher. Number of safety shut-downs was 11. The reasons were related to electrical faults, 9 occurrences, failure of the instrument for reactor period measurement, once, and failure of the pump once. This report covers state and operation data of the reactor components, control instruments, devices for reactor protection, dosimetry, heavy water system, helium system, mechanical and electrical equipment. Problems related to training of the staff and lack of personnel are mentioned as well.

  5. RA Research reactor, Annual report 1971; Istrazivacki nuklearni reaktor RA - Izvestaj za 1971. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1971-12-15

    effects. In the introduction of this report it has been emphasised that the decision makers should have in mind the negative effects of low budget on the reactor safe and reliable operation. For the sake of reactor, decision about the future operation and financing should be done as soon as possible, either to cease operation or continue with adequate financial support. [Serbo-Croat] Reaktor RA je u 1971. godini radio na nominalnoj snazi 190 dana i 50 dana na manjim snagama. Ukupni rad iznosio je 31606 MWh odnosno 5,3% vise od planiranog, sto je najvisa vrednost od kako je reaktor pusten u rad. Reaktor je koriscen za ozracivanja i eksperimente za 425 korisnika od cega 370 iz Instituta i 55 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Odstupanja od plana, odnosno veceg ostvarenog rada bilo je u junu i decembru usled posebnih zahteva korisnika. Ukupni broj prekida rada bio je manji od svih prethodnih godina, uglavnom zbog manjeg broja nestanka napona u vreme rada reaktora. U toku godine bilo je samo jedno sigurnosno zaustavljanje, ciji je uzrok bila pojava laznog signala opreme za zastitu reaktora. Nijednog duzeg prekida rada nije bilo zbog neispravnosti opreme. Kracih prekida bilo je usled kidanja spojki na potisnom cevovodu tehnicke vode, sto je bilo izazvano klizanjem zemljista u podrucju crpne stanice na Dunavu. Ukupna doza ozracivanja ljudstva bila je manja nego prethodnih godina. Nije bilo ni jednog akcidenta niti slucaja koji bi se mogao nazvati akcidentom. Dekontaminirano je znatno manje povrsina nego ranijih godina. Zakljuceno je da je uspesan rad reaktora u 1971. godini rezultat valjanog rada u prethodnim godinama. Medjutim usled jos nedefinisane politike u pogledu buduceg rada, odnosno neizvesnosti u vezi finansiranja, neki poslovi su obustavljeni. Tu spada proucavanje mogucnosti prelaska na koriscenje visokoobogacenog goriva sto bi povecalo korisni neutronski fluks i ucinilo reakor konkurentnim za

  6. Nuclear Reactor RA Safety Report, Vol. 5, Reactor cooling systems; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 5, Hladjenje reaktora i pripadajuci sistemi

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    RA reactor cooling system enable cooling during normal operation and under possible accidental conditions and include: technical water system, heavy water system, helium gas system, system for heavy water purification and emergency cooling system. Primary cooling system is a closed heavy water circulation system. Heavy water system is designed to enable permanent circulation and twofold function of heavy water. In the upward direction of cooling it has a coolant role and in the downward direction it is the moderator. Separate part of the primary coolant loop is the system for heavy water purification. This system uses distillation and ion exchange processes. [Serbo-Croat] Sistemi za hladjenje reaktora RA obezbedjuju hladjenje u svim rezimima eksploatacije i potenijalno udesnim satnjim i obuhvataju: sistem tehnicke vode, sistem teske vode, gasni sistem helijuma, sistem za preciscavanje teske vode i sistem za akcidentalno hladjenje. Primarni sistem za hladjenje je zatvoreni cirkulacioni sistem teske vode. Specificnost sistema teske vode vezana je za neprekidnost cirkulacije i razlicite funkcije teske vode. U uzlaznom strujanju, teske vode ima ulogu hladioca a u silaznom ulogu moderatora. Poseban deo primarnog sistema predstavlja sistem za preciscavanje teske vode. Ovaj sistem koristi destilacioni i jonoizmenjivacki postupak.

  7. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Dixon, D.; Lord, D.

    1978-01-01

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

  8. Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

    International Nuclear Information System (INIS)

    Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

    2017-01-01

    Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

  9. Selective Insulin Resistance and the Development of Cardiovascular Diseases in Diabetes: The 2015 Edwin Bierman Award Lecture

    Science.gov (United States)

    Park, Kyoungmin; Li, Qian

    2016-01-01

    The Edwin Bierman Award Lecture is presented in honor of the memory of Edwin L. Bierman, MD, an exemplary scientist, mentor, and leader in the field of diabetes, obesity, hyperlipidemia, and atherosclerosis. The award and lecture recognizes a leading scientist in the field of macrovascular complications and contributing risk factors in diabetes. George L. King, MD, of the Section of Vascular Cell Biology and Complications, Dianne Nunnally Hoppes Laboratory for Diabetes Complications, Joslin Diabetes Center, Harvard Medical School, Boston, MA, received the prestigious award at the American Diabetes Association’s 75th Scientific Sessions, 5–9 June 2015, in Boston, MA. He presented the Edwin Bierman Award Lecture, “Selective Insulin Resistance and the Development of Cardiovascular Disease in Diabetes,” on Sunday, 7 June 2015. This review is focused on the factors and potential mechanisms that are causing various cardiovascular pathologies. In diabetes, insulin’s actions on the endothelium and other vascular cells have significant influence on systemic metabolisms and the development of cardiovascular pathologies. Our studies showed that insulin receptors on the endothelium are important for insulin transport across the endothelial barrier and mediate insulin’s actions in muscle, heart, fat, and the brain. Insulin actions on the vascular cells are mediated by two pathways involving the actions of either IRS/PI3K/Akt or Grb/Shc/MAPK. Insulin’s activation of IRS/PI3K/Akt results in mostly antiatherogenic actions, as this pathway induces activation of eNOS, the expressions of HO-1 and VEGF, and the reduction of VCAM-1. In contrast, insulin’s activation of the Grb/Shc/MAPK pathway mediates the expressions of ET-1 and PAI-1 and migration and proliferation of contractile cells, which have proatherogenic actions. Elevated levels of glucose, free fatty acids, and inflammatory cytokines due to diabetes and insulin resistance selectively inhibit insulin

  10. Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

    International Nuclear Information System (INIS)

    Haferkamp, D.

    2008-01-01

    Between 1994 and 1998 the equipment for dry reloading of a research reactor was developed by Noell, which was funded by the German Federal Government and State of Saxonia. The task of this development programme was the design and delivery of an equipment able to load the spent fuel into the shipping casks in a dry mode for research reactors, where wet loading inside the storage pool is impossible. ALARA and infrastructure conditions had to be taken into consideration. Most of the research reactors of TRIGA MARK I type or WWR-SM have operating modes for handling of spent fuel inside the pond or for transfer of spent fuel from pond to dry/wet storage pools. On the other hand, most of them cannot handle heavy weighted shipping casks inside the reactor building because of the crane capacity, or inside water pool because of dimensions and weight of shipping casks. A typical licensed normal operating procedure for spent fuel in research reactors (TRIGA MARK I) is shown. Dry unloading procedure is described. Additionally to the normal operating procedures at the MHH research reactor the following steps were necessary: - dry packaging of spent fuel elements into the loading units (six packs) in order to minimise the transfer and loading steps between the pool and shipping cask; - transfer of spent fuel loading units from dry storage pool to the shipping cask (outside the reactor building) in a shielded transfer cask; - dry reloading of loading units, into the shipping casks outside the reactor building. The Dry Reloading Equipment implies the following 5 items: 1. loading units (six packs), which includes: - capacity up to six spent fuel elements; - criticality safe placement of spent fuel elements; - handling of several spent fuel elements in an aluminium loading unit. 2. Special Transfer Cask, which includes: - shielded housing with locks; - gripper inside housing; - hoist outside housing; - computer aided operation mode for loading and unloading. 3. Transfer Vehicle

  11. The TEX-I real-time expert system, applied to situation assessment for the SNR-300 reactor

    International Nuclear Information System (INIS)

    Schmal, N.; Leder, H.J.; Schade, H.J.

    1988-01-01

    Interatom, a subsidiary company of Siemens, is developing expert systems for the technical domain. These systems are operating in various industrial applications like flexible manufacturing or plant configuration, based on a domain-specific expert system shell, developed by Interatom. Additional projects are focusing on real-time diagnostics, e.g., for nuclear power plants. The authors report in this paper about a diagnosis expert system for the liquid-metal fast breeder reactor SNR-300, which uses new real-time tools, developed within the German TEX-I project (technical expert systems for data interpretation, diagnosis, and process control). The purpose of the system is to support the reactor operators in assessing plant status in real time, based on readings from many sensors. By on-line connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components, and recommended operator response in abnormal conditions. In the present knowledge acquisition and test phase, the expert system is connected to a real-time simulation of the reactor. The simulator is based on a thermohydraulic code for simulation of the transient behavior of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers, and cooling components. Additionally, the system's response to an asynchronous operator interaction can be simulated

  12. Stationary liquid fuel fast reactor SLFFR – Part I: Core design

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

  13. Biologist Edwin Grant Conklin and the idea of the religious direction of human evolution in the early 1920s.

    Science.gov (United States)

    Pavuk, Alexander

    2017-01-01

    Edwin Grant Conklin, renowned US embryologist and evolutionary popularizer, publicly advocated a social vision of evolution that intertwined science and modernist Protestant theology in the early 1920s. The moral prestige of professional science in American culture - along with Conklin's own elite scientific status - diverted attention from the frequency with which his work crossed boundaries between natural science, religion and philosophy. Writing for broad audiences, Conklin was one of the most significant of the religious and modernist biological scientists whose rhetoric went well beyond simply claiming that certain kinds of religion were amenable to evolutionary science; he instead incorporated religion itself into evolution's broadest workings. A sampling of Conklin's widely-resonant discourse suggests that there was substantially more to the religion-evolution story in the 1920s US than many creationist-centred narratives of the era imply.

  14. Program of critical experiment and measurements at the RA reactor; Program kriticnih eksperimenata i merenja na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-14

    Program described in this document describes in detail the following experiments: critical experiments with two reactor core lattices with 38 and 44 fuel channels, initial heavy water level being 1300 mm, criticality is achieved by adding heavy water; preliminary analysis of heavy water quality and verification of the fuel isotopic contents; experiment with the initial core which contains 56 fuel channels with maximum heavy water level according to the Russian proposal; measurement of neutron flux by Dy and In foils; measurement of reactivity excess dependent on the heavy water level and number of fuel rods; measurement of reactor period for determined reactivity change; measurement of moderator temperature coefficient; measurement of absolute flux. [Serbo-Croat] Program sadrzan u ovom dokumentu opisuje detaljno sledece eksperimente: kriticni eksperiment sa dve konfiguracije jezgra reaktora, sa 38 i 44 gorivna kanala, pocetni nivo teske vode je 1300 mm, kriticnost se dostize dodavanjem teske vode; prethodno izvrsenom analizom teske vode i proverom izotopskog sastava goriva; eksperiment sa pocetnom resetkom koja prema ruskom predlogu sadrzi 56 gorivnih kanala i maksimalnom visinom teske vode; merenje raspodele neutronskog fluksa folijama Dy i In; kalibracija regulacionih sipki; merenje viska reaktivnosti sa promenom visine nivoa teske vode i promenom broja sipki; merenje periode reaktora za odredjenu promenu reaktivnosti; merenje temperaturnog koeficijenta za vodu; merenje apsolutnog fluksa.

  15. RA Reactor operation and maintenance (I-IX), part VII, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora (I-IX), VII Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    This volume covers the following reports concerned with the maintenance and repair work of the RA reactor: repair of the technical water system; maintenance of the transportation equipment; vacuuming and drying during refurbishment; repair and decontamination of the distillation device; and the report on participation of the operational dosimetry division in the RA reactor refurbishment activities.

  16. RA Reactor operation and maintenance (I-IX), part VI, Task 3.08/04, Refurbishment of the RA reactor; Pogon i odrzavanje reaktora (I-IX), VI Deo, Zadatak 3.08/04 Remont reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    During the period planned for maintenance and refurbishment of the RA reactor the gas reactor system including the ventilation system was inspected and tested, the components were cleaned. This report describes detailed instructions and actions concerning repair and decontamination of the gas and ventilation systems components.

  17. RA reactor operation and maintenance, Annual report 1974; Pogon i odrzavanje reaktora RA - Izvestaj o radu u 1974. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1974-12-15

    During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m{sup 2} of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems. [Serbo-Croat] Reaktor RA je u 1974. godini radio na nominalnoj snazi 194 dana i 13 dana na manjim snagama. Ukupni rad iznosio je 30711 MWh odnosno 1,4% vise od planiranog. Prakticno nije bilo odstupanja od plana rada. Reaktor je koriscen za ozracivanja i eksperimente za 437 korisnika. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. U toku godine bilo je 11 sigurnosnih zaustavljanja, od cega 8 zbog elektricnog napona i 3 usled greske osoblja. Ukupna doza ozracivanja ljudstva bila je manja nego prethodnih godina. Maksimalna doza po oveku bila je 50% manja od maksimalno dozvoljene doze. Nije bilo ni jednog akcidenta. Dekontaminirano je znatno manje povrsina nego ranijih godina, i sakupljeno manje otpada nego prethodnih godina, dok tecnih efluenata nije bilo. Zakljuceno je da uspesan rad reaktora u 1974. godini ima poseban znacaj kada se imaju na umu problemi finansiranja reaktora.

  18. Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

    Science.gov (United States)

    Carré, Frank

    2014-09-01

    Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  19. Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

    Directory of Open Access Journals (Sweden)

    Carré Frank

    2014-01-01

    Full Text Available Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

  20. Higher plant availability and reduced reactor scram frequency in PWRs by appropriate system and I and C design

    International Nuclear Information System (INIS)

    Frei, G.; Weber, J.

    1987-01-01

    High plant availability and reliability are guaranteed by appropriate design of reactor and BOP systems, this including the plant I and C systems. It is of advantage to have design, construction and commissioning of the plant concentrated in the hands of a single company to avoid interface problems between the different areas of the plant. The integrated overall control concept developed by KWU with control, limitation and protection systems as well as optimized operational and monitoring systems assisted by instrumentation channel redundance and logic for selection of the second highest (or second lowest) signal value as appropriate for comparison with limitation setpoints, minimize the severity of transients. This results in a reduction in the frequency of reactor scrams and of unnecessary actuation of safety systems. Dynamic plant behavior is described for a number of examples where the improved plant behavior resulting from the above design features enhances plant availability

  1. RA Research reactor, Annual report 1970 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1970. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1970-12-15

    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

  2. Radiation effects and tritium technology for fusion reactors. Volume I. Proceedings of the international conference, Gatlinburg, Tennessee, October 1--3, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Watson, J.S.; Wiffen, F.W.; Bishop, J.L.; Breeden, B.K. (eds.)

    1976-03-01

    Separate abstracts were prepared for the 29 included papers in Vol. I. The topics covered in this volume include swelling and microstructures in thermonuclear reactor materials. Some papers on modeling and damage analysis are included. (MOW)

  3. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  4. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    International Nuclear Information System (INIS)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-01

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor

  5. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-15

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

  6. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1977-01-01

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  7. Investigation of Equilibrium Core by recycling MA and LLFP in fast reactor cycle (I)

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Shono, Akira; Ishikawa, Makoto

    1999-05-01

    Feasibility study on a self-consistent fuel cycle system is performed in the nuclear fuel recycle system with fast reactors. In this system, the self-generated MAs (Minor Actinides) and LLFPs (Long Lived Fission Products) are confined and incinerated in the fast reactor. Analyses of the nuclear properties for an 'Equilibrium Core', in which the self-generated MAs and LLFPs are confined, are investigated. A conventional sodium cooled oxide fuel fast reactor is selected as the core specifications for the 'Equilibrium Core'. This 600 MWe fast reactor does not have a radial blanket. In this study, the nuclear characteristics of the 'Equilibrium Core' are compared with those of a 'Standard Core' and '5 w/oMA Core'. The 'Standard Core' does not confine MAs and LLFPs in the core, and a 5 w/o-MA Rom LWR is loaded in the '5 w/oMA Core'. Through this comparison between 'Equilibrium Core' and the others, the specific characters of the 'Equilibrium Core' are investigated. In order to realize the 'Equilibrium Core' in the viewpoint of nuclear properties, whether the conventional design concept of fast reactors must be changed or not is also evaluated. The analyses for the nitride and metallic fuel cores are also performed because of their different nuclear characteristics compared with the oxide fuel core. Assuming the separation of REs (Rare Earth elements) from MAs and the isotope separation of LLFPs, most of the nuclear properties for the 'Equilibrium Core' are not beyond those for the '5 w/oMA Core'. It is, therefore, possible to bring the 'Equilibrium Core' into existence without any drastic modification for the design concept of the typical oxide fuel fast reactors. Although the 15.1[w/o] LLFPs are loading in the core of the oxide fuel 'Equilibrium Core', a breeding ratio is more than 1.0 and the difference in a amount of plutonium between a charging and discharging is only 0.04 [ton/year]. Without any drastic change for the design concept of the conventional oxide fuel

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-15

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

  10. RA Research reactor, Annual report 1968 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1968. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    During 1968, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and during 50 days at lower power levels. Total production amounted to 31051 MWh which is 3.5% higher than planned. reactor was used for irradiation and experiments according to the demand of 600 users, of which 517 from the Institute and 83 externals users. This report contains detailed data about reactor power and experiments performed in 1968. It is concluded that the reactor operation was more successful than during previous years. There was only one longer interruption which lasted 27 hours because of the power cut on the cable for the pump station on Danube. Number of safety shutdowns were at the same level as during last year. The only significant incident in 1968 was air contamination with the radioactive argon in the reactor hall. The reactor operation was not interrupted although the hall was evacuated for two hours. The was no significant exposure of the staff. In April and September the integral dosed were higher than during other months because of the accident during refueling (mixing the slugs with irradiated and fresh fuel). There was no significant surface contamination, i.e. the decontaminated surface were negligible. Due to 'mixing' refueling scheme. [Serbo-Croat] Reaktor RA je u 1968. godini radio na nominalnoj snazi od 6,5 MW 190 dana i 50 dana na manjim snagama. Ukupni rad iznosio je 31051 MWh odnosno 3,5% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 600 korisnika od cega 517 iz Instituta i 83 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Zakljucuje se da je reaktor radio uspesnije nego prethodnih godina. U toku 1968. godine samo je jedan duzi prekid u radu od 27 casova izazvan zbog proboja kablovske glave na odvodu za pumpnu stanicu na Dunavu. Sigurnosna zaustavljanja bila su na proslogodisnjem nivou. Jedini znacajniji incident u 1968. godini, bio je kontaminacija vazduha

  11. Flow effect on {sup 135}I and {sup 135}Xe evolution behavior in a molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Jianhui; Guo, Chen [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Cai, Xiangzhou [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Yu, Chenggang; Zou, Chunyan [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Han, Jianlong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jingen, E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China)

    2017-04-01

    Highlights: • {sup 135}Xe and {sup 135}I evolution law in a molten salt reactor is analytically deduced. • The circulation of fuel salt through the primary loop decreases the concentration of {sup 135}I and {sup 135}Xe. • {sup 135}I and {sup 135}Xe concentration reduction is independent with the mass flow rate at normal core operating condition. • Increasing the external core volume would raise {sup 135}I and {sup 135}Xe concentration reduction caused by the flow effect. - Abstract: Molten Salt Reactor (MSR) employs fissile material dissolved in the fluoride salt as fuel which continuously circulates through the primary loop with the flow cycle time being a few tens of seconds. The nuclei evolution law is quite different from that in a solid fuel reactor. In this paper, we analytically deduce the nuclei evolution law of {sup 135}Xe and {sup 135}I which are entrained in the flowing salt, evaluate its concentration changing with the burnup time, and validate the result with the SCALE6. The circulation of fuel salt could decrease the concentration of {sup 135}Xe and {sup 135}I, and the reduction can achieve to around 40% and 50% for {sup 135}Xe and {sup 135}I respectively at a small power level (e.g., 2 MW) when the core has the same fuel salt volume as that of the outer-loop. Furthermore, it can be found that the reduction is inversely proportional to the core to outer-loop volume ratio, but uncorrelated with the mass flow rate under normal operating condition of a MSR. At low core power scale, the flow effect on {sup 135}Xe concentration reduction is apparent, but it is mitigated as the core power scale increases because of the rise of {sup 135}I concentration, which raises its decay to {sup 135}Xe and compensates the loss of {sup 135}Xe due to decay at the outer-loop. The decreased {sup 135}Xe concentration results in a core reactivity increase varying from around 150 pcm to 1000 pcm depending on the core power and core to outer-loop volume ratio.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-12-15

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

  13. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1962-09-15

    In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined.

  14. Nuclear data for reactors. Proceedings of the second international conference. Vol. I

    International Nuclear Information System (INIS)

    1970-01-01

    The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

  15. Operation and maintenance of the RA reactor in 1965; Pogon i odrzavanje reaktora RA. Izvestaj o radu u 1965. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-01-15

    It has been planned for 1965 that the RA reactor would be operated each month for 20 days at nominal power of 6.5 MW, at lower power for 5 days, meaning production of 27 400 MWh. The plan was fulfilled since reactor produced 28809 MWh, i.e. 5% more than planned. Reactor was used for irradiation in the vertical experimental channels according to the demand of 1264 users from the Institute and 191 external users. Two groups of experiments done: at nominal power simultaneously with isotope production and experiments which demanded particular power levels and temperatures. Three fuel exchanges were done during this year, meaning that 40 fuel channels were changed in total. Vertical experimental channels VEK-1 and VEK-9 having diameter 100 mm were changed by channels having diameter 50 mm and shortened by 435 mm. Channel VEK-5 with diameter 110 mm was changed shortened by 430 mm. This enabled better fuel economy, the burnup was increased from 4500 MWd/t to 5000 MWd/t. This report contains the action plan for 1966.

  16. Costs of magnets for large fusion power reactors: Phase I, cost of superconductors for dc magnets

    International Nuclear Information System (INIS)

    Powell, J.R.

    1972-01-01

    Projections are made for dc magnet conductor costs for large fusion power reactors. A mature fusion economy is assumed sometime after 2000 A. D. in which approximately 90,000 MW(e) of fusion reactors are constructed/year. State of the art critical current vs. field characteristics for superconductors are used in these projections. Present processing techniques are used as a basis for the design of large plants sized to produce approximately one-half of the conductor needed for the fusion magnets. Multifilamentary Nb-Ti, Pb-Bi in glass fiber, GE Nb 3 Sn tape, Linde plasma sprayed Nb 3 Sn tape, and V 3 Ga tape superconductors are investigated, together with high purity aluminum cryoconductor. Conductor costs include processing costs [capital (equipment plus buildings), labor, and operating] and materials costs. Conductor costs are compared for two sets of material costs: current (1971 A. D.) costs, and projected (after 2000 A. D.) costs. (U.S.)

  17. HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

    International Nuclear Information System (INIS)

    Schultz, J.H.; Montgomery, D.B.

    1983-04-01

    HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

  18. Light-water reactors: preliminary safety and environmental information document. Volume I

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning the reference PWR reactor system; once-through, low-enrichment uranium-235 fuel, 30 MWD per kilogram (PWR LEU(5)-OT); once-through, low-enrichment, high-burnup uranium fuel (PWR LEU(5)-Mod OT); self-generated plutonium spiked recycle (PWR LEU(5)-Pu-Spiked Recycle); denatured uranium-233/thorium cycle (PWR DU(3)-Th Recycle DU(3)); and plutonium/thorium cycle

  19. Regulations and instructions for RA reactor operation; Propisi i uputstva za pogon reaktora 'A'

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-07-01

    This regulatory guide consists of following 4 chapters: Description of the RA reactor, organization scheme, regulations for performing experiments; Regulations for staff on duty; Instructions for operating the vacuum systems, heavy water and helium systems; and evacuation in case of accident. [Serbo-Croat] Ovaj pravilnik sadrzi sledeca 4 poglavlja: Opis reaktora RA, sema organizacije rada, propisi za izvodjenje eksperimenata; Pravilnik za rad dezurnog osoblja; Uputstva za rada sa vakuum sistemima, sistemom teske vode, sistemom helijuma; evakuacija u slucaju udesa.

  20. Analysis and primary design of the I and C system architecture for HTGR-10 reactor

    International Nuclear Information System (INIS)

    Yang Zijue; Zhao Guoji

    1993-01-01

    The consideration of making good use of the-state-of-the-arts technology in designing advanced Instrumentation and Control System architecture is discussed. A fully distributed and fully micro-computerized, local network based Instrumentation and Control System is designed for the HTR-10 reactor. The advantages of the system architecture include high reliability and availability, flexibility, economics, etc. It also fits for other production processes

  1. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  2. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  3. Development of database system on MOX fuel for water reactors (I)

    International Nuclear Information System (INIS)

    Kikuchi, Keiichi; Nakazawa, Hiroaki; Abe, Tomoyuki; Shirai, Takao

    2000-04-01

    JNC has been conducted a great number of irradiation tests to develop MOX fuels for Advanced Thermal Reactor and Light Water Reactors. In order to manage irradiation data consistently and to effectively utilize valuable data obtained from the irradiation tests, we commenced construction of database system on MOX fuel for water reactors in 1998 JFY. Collection and selection of irradiation data and relevant fuel fabrication data, design of the database system and preparation of assisting programs have been finished and data registration onto the system is under way according to priority at present. The database system can be operated through the menu screen on PC. About 94,000 records of data on 11 fuel assemblies in total have been registered onto the database up to the present. By conducting registration of the remaining data and some modification of the system, if necessary, the database system is expected to complete in 2000 JFY. The completed database system is to be distributed to relevant sections in JNC by means of CD-R as a media. This report is an interim report covering 1998 and 1999 JFY, which gives the structure explanation and users manual concerning to the prepared database up to the present. (author)

  4. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  5. Dosimetry and radiation protection at the RA reactor in 1972; Dozimetrija i zastita kod reaktora RA u 1972. godini

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1973-07-01

    Dosimetry results collected within radiation protection of the RA reactor during this year are presented. Neutron and gamma radiation data were measured at characteristic control points. Statistical review of the total number of measurement is given as well. The report includes contents of radioactive gasses and aerosols in the air, as well as the contamination data of surfaces, clothes and uncovered body parts of the personnel. Particular accident which occurred during dismantling of the experimental channel containing the capsule with the new fuel element was analysed. This accident occurred at the RA reactor at the beginning of this year. It was found that that the maximum individual external dose was 2.2 R, and that only one individual was exposed to this dose. About 15% of the personnel was exposed to doses between 1 and 2 R, the remaining 85% was exposed to doses less than 1 R. Base on the frequency of activities undertaken in the contaminated regions, safety and control measures and expected internal exposure of the personnel, it was evaluated that the internal exposure could be neglected compared to the external exposure of the personnel. Prikazani su rezultati sakupljani u toku godine u okviru dozimetrijske kontrole i zastite kod reaktora. Dati su podaci o nivoima neutrona i gama zracenja na karakteristicnim kontrolnim mestima, kao i statisticki pregledi ukupnog broja merenja. Navedeni su rezultati merenja sadrzaja radioaktivnih gasova i aerosola u vazduhu, kao i stepena kontaminacije povrsina, odece i otkrivenih delova tela radnog osoblja. Analiziran je specifican akcident koji se odigrao na reaktoru pocetkom godine, pri demontazi eksperimentalnog kanala sa kapsulom novog gorivog elementa. Na kraju, izlozena je analiza ozracivanja radnog osoblja. Konstatovano je da je maksimalna individualna doza spoljasnjeg ozracivanja bila 2,2 (R), i da je ovoj dozi bilo izlozeno samo jedno lice. Oko 15% osoblja bilo je izlozeno dozama izmedju 1 i 2 (R), a ostalih 85

  6. RA Research reactor, Annual report 1969; Istrazivacki nuklearni reaktor RA - Izvestaj za 1969. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1969-12-15

    During 1969, the RA Reactor was operated at nominal power of 6.5 MW for 200 days, and 15 days at lower power levels. Total production mounted to 31131 MWh which is 3.77% higher than planned. Reactor was used for irradiation and experiments according to the demand of 463 users from the Institute and 63 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. If there had been no problems with power supply during last three months and Danube low water level in September and October the past year would have been the most successful up to now. The number od scram shutdowns was not higher than during past two years in spite of the difficulties in the last quarter. There were three incidents which caused higher personnel exposure during operation. One, was the destruction of the canner with silver (because the time spent in the core was too long) which caused the surface contamination of the platform, the background radiation was 10 to 100 times higher than regular. The other two cases were caused by failure of the device for handling the fuel slugs in the fuel channels during refuelling. Reactor refuelling was done four times during 1969, and 499 fresh fuel slugs were used. Refuelling applied the approach of 'mixing' the fresh fuel slugs with the 'old' fuel slugs in the fuel channel. Decontamination of surfaces was on the same level as previously in spite of the problems with silver. Since two staff members have left, the present number od employees is now the minimum needed for reactor operation and maintenance. It is stated that the operation of components and equipment is on sufficiently high level after ten years of reactor operation. The action plan for 1970 is made according to the same principles as in previous four years but the planned production is decreased to 25000 MWh, because control of important components is needed after ten

  7. Independent CO{sub 2} loop for cooling the samples irradiated in vertical experimental channels of the RA reactor, Vol. I; Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA, Album I

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-06-15

    Project 'independent CO{sub 2} loop for cooling the samples irradiated in the vertical experimental channels of the RA reactor' is presented in two volumes: volume I - head of the low temperature coolant loop for reactor RA, and volume II - Outer low-temperature reactor coolant loop. Volume I includes: the design specifications for the head of the low-temperature coolant loop, technical description, thermal calculation, calculations of mechanical loads, antireactivity and activation of the components of the coolant loop head, engineering schemes and drawings, cost estimation data. [Serbo-Croat] Projekat 'Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA', sastoji se od dva albuma: album I - Glava niskotemperaturno rashladne petlje za reaktor RA, album II - Spoljno kolo niskotemperaturne rashladne petlje za reaktora. Album I sadrzi projektni zadatak glave niskotemperaturne petlje, tehnicki opis, termicki proracun, proracun mehanickih naprezanja, antireaktivnosti i aktivacije kontrukcionih elemenata glave petlje, konstrukcione seme i crteze glave petlje, predracun.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-15

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

  9. RB Research nuclear reactor, Annual report for 1995, I-IV; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1995. godini, I-IV

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-15

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor.

  10. Initiation of a phase-I trial of neutron capture therapy at the MIT research reactor

    International Nuclear Information System (INIS)

    Harling, O.K.; Bernard, J.A.; Yam, Chun-Shan

    1995-01-01

    The Massachusetts Institute of Technology (MIT), the New England Medical Center (NEMC), and Boston University Medical Center (BUMC) initiated a phase-1 trial of boron neutron capture therapy (BNCT) on September 6, 1994, at the 5-MW(thermal) MIT research reactor (MITR). A novel form of experimental cancer therapy, BNCT is being developed for certain types of highly malignant brain tumors such as glioblastoma and melanoma. The results of the phase-1 trials on patients with tumors in the legs or feet are described

  11. Coupled map lattice (CML) approach to power reactor dynamics (I) - preservation of normality

    International Nuclear Information System (INIS)

    Konno, H.

    1996-01-01

    An application of coupled map lattice (CML) model for simulating power fluctuations in nuclear power reactors is presented. (1) Preservation of Gaussianity in the point model is studied in a chaotic force driven Langevin equation in conjunction with the Gaussian-white noise driven Langevin equation. (2) Preservation of Guassianity is also studied in the space-dependent model with the use of a CML model near the onset of the Hopf bifurcation point. It is shown that the spatial dimensionality decreases as the maximum eigenvalue of the system increases. The result is consistent with the observation of neutron fluctuation in a BWR. (author)

  12. Automated systems help prevent operator error during [reactor] I and C [instrumentation and control] testing

    International Nuclear Information System (INIS)

    Courcoux, R.

    1989-01-01

    On a nuclear steam supply system, even a minor failure can involve actuation of the whole reactor protection system (RPS). To reduce the likelihood of human error leading to unwanted trips during the maintenance of instrumentation and control systems, Framatome has been developing and installing various automated testing systems. Such automated systems are particularly helpful when periodic tests with a potential for RPS actuation have to be carried out, or when the test is on the critical path for the refuelling outage. The Sensitive Channel Programme described is an example of the sort of work that has been done. (author)

  13. RA Research reactor, Annual report 1972; Istrazivacki nuklearni reaktor RA - Izvestaj za 1972. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1972-12-15

    During 1972, the total production was 31151 MWh which is 3.8% higher than planned. The reactor was used for irradiation and experiments according to the demand of 381 users, of which 340 from the Institute and 41 external users. This report contains detailed data about reactor power and experiments performed in 1972. Discrepancies from the action plan, meaning higher production was achieved due to special demands of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to announced power cuts. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. There were no longer interruptions. One shorter interruption (shorter than 24 hours) caused by removal of a UO{sub 2} capsule from the core, placed there for measuring heat transfer. Total personnel exposure dose was lower than during previous years. One accident caused contamination with gases and aerosols containing mainly shot-living isotopes. Decontamination od surfaces was less than during previous years. Practically there was no surface contamination that would demand action of the decontamination team, except for the regular decontamination after refueling. It was concluded that the successful operation in 1972 has a special significance having taking in account the financial crisis caused by the unresolved status of the reactor. It is emphasised, in the plan for the next year that there is an urgent need of making a long-term plan of rector application. It is indispensable to finish preparatory tasks for replacing the fuel with the highly enriched fuel elements by 1974, and building the core emergency cooling system. [Serbo-Croat] Ukupni rad Reaktora RA je u 1972. godini iznosio je 31151 MWh odnosno 3,8% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 381 korisnika od cega 340 iz Instituta i 41 za korisnike izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i

  14. Modification of OCA-I for application to a reactor pressure vessel with cladding on the inner surface

    International Nuclear Information System (INIS)

    Sauter, A.; Cheverton, R.D.; Iskander, S.K.

    1983-01-01

    The computer code OCA-I calculates the temperature distribution through the walls of a cylinder during a thermal transient and then performs a two-dimensional linear-elastic fracture-mechanics analysis to obtain stress-intensity factors for long surface flaws, considering both pressure and thermal loads. The code has been particularly useful in evaluating flaw behavior in reactor pressure vessels during overcooling accidents, but it has not previously treated the stainless steel cladding on the inner surface of the vessel as a discrete region. Although the cladding is quite thin compared with the base material, the large difference in thermal conductivity and coefficient of thermal expansion between the two materials results in a significant effect of the cladding on stress-intensity factors for surface cracks. Thus, the cladding was recently included as a discrete region in OCA-I

  15. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Huda, M.Q.; Chakrobortty, T.K.; Rahman, M.; Sarker, M.M.; Mahmood, M.S.

    2003-05-01

    This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

  16. Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1979-05-01

    The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

  17. Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

    International Nuclear Information System (INIS)

    1979-05-01

    The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

  18. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

  19. The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

    Directory of Open Access Journals (Sweden)

    LI U Ja

    2017-02-01

    Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations,the fi- nite element coupling in the transient electromagnetic field coupling,structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution,core magnetostrictive displacement,max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor,the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated,the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment,it is proved multi-physics coupling is an effective method for forecast of noise.

  20. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    International Nuclear Information System (INIS)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  1. Acoustic emission during the elastic-plastic deformation of low alloy reactor pressure vessel steels. I

    International Nuclear Information System (INIS)

    Holt, J.; Goddard, D.J.

    1980-01-01

    Measurements of the acoustic emission behaviour of A533B and C-Mn low alloy reactor pressure vessel steels subjected to uniaxial tensile deformation are described. The effects on the emission activity of the rolling plane orientation and the carbide morphology were examined. Detailed discussions are given of the stress dependence of the emission activity below yield and of its recovery by annealing at the stress relief temperature. It is shown that the dominant emission source is the same in both steels and is associated with inclusions, such as MnS, elongated by the rolling process, the carbide morphology being relatively unimportant. A criterion for the occurrence of an emission is obtained which is directly analogous to the general criterion for yielding. It is also shown that a large fraction, at least, of the emission activity arises from a recoverable process such as localized yielding around inclusions or limited inclusion decohesion and not from inclusion fracture. Low activity in C-Mn steel taken from reactor pressure vessels, previously attributed to spheroidization of carbides, is shown to be due to the limited acoustic recovery of these relatively high sulphur content steels when annealed at the stress relief temperature. It is concluded that the limited amplitudes of these emissions during deformation severely restrict their potential application in practice. (Auth.)

  2. Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

    International Nuclear Information System (INIS)

    Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho; Son, Han Seong; Song, Deok Yong

    2007-01-01

    As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code

  3. Fourteenth Annual Meeting of the International Working Group on Past Reactors. Summary Report. Part I

    International Nuclear Information System (INIS)

    1981-11-01

    The Fourteenth Annual Meeting of the IAEA-IWGFR was held in accordance with the recommendations of the previous Annual Group Meeting, at the Vienna International Centre, Vienna from 31 March to 3 April 1981. All Member States of the group were represented at the meeting: France, the Federal Republic of Germany, India, Italy, Japan, the Union of Soviet Socialist Republics, the United Kingdom and the United States of America. The meeting was also attended by representatives from the Commission of European Communities, the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development, the International Atomic Energy Agency and observers from Switzerland. The Agenda of the Meeting was as follows: 1. Review of IWGFR activities; 2. Consideration of future method of operation of the IWGFR; 3. Consideration of Conferences on Fast Reactors; 4. Consideration of the major recommendations of some of the IWGFR specialists' meetings for which the support of the IWGFR is requested; 5. Consideration of a schedule for specialists' meetings in 1981-1982; 6. Presentations and discussions on national programmes on fast breeder reactors.; 7. Recommendation of the IWGFR regarding a request of Switzerland concerning participation in the IWGFR; 8. The date and place of the Fifteenth Annual Meeting of the IWGFR

  4. Hatch 1:100000 Quad Hydrography DLGs

    Data.gov (United States)

    Earth Data Analysis Center, University of New Mexico — Digital line graph (DLG) data are digital representations of cartographic information. DLG's of map features are converted to digital form from maps and related...

  5. ARM tõestab, et ka analüütikud eksivad / Fredy-Edwin

    Index Scriptorium Estoniae

    Esse, Fredy-Edwin

    2011-01-01

    Apple'i iPadidele ja iPhone'idele kiipe tootva Suurbritannia firma ARM Holdings aktsia on kõige suurem tõusja Londoni FTSE 100 indeksi kuue kuu, 12 kuu, 18 kuu, kahe aasta ja kolme aasta lõikes, samas on analüütikud kõigist indeksi aktsiatest kõige negatiivsemalt meelestatud just ARM Holdingsi suhtes. Graafik

  6. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

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

  7. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  8. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1986. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues. [Serbo-Croat] Sa ciljem da se obezbedi pouzdan rad reaktora RA a u skladu sa zakonskim propisima, zavrsena su tri velika zahvata zapoceta 1984: izgradnja novog sistema za udesno hladjenje, rekonstrukcija postojeceg sistema za ventilaciju, i modernizacija reaktorske instrumentacije. Istovremeno tokom 1985/1986. zapoceta je modernizacija instrumentacije i rekonstrukcija sistema za rukovanje i skladistenje iskoriscenog goriva u zgradi reaktora. Projekti za navedene radove su vec zavrseni ili su u zavrsnoj fazi, a ocekuje se da ce rekonstrukcija oba sistema biti zavrsena do kraja 1988. odnosno sredine 1989. godine. Izrada izvestaja o sigurnosti reaktora RA, prema preporukama MAAE zavrsena je 1986. Investiciona ulaganja na reaktoru Ra u 1986. iskoriscena su za: nabavku 8000 kg teske vode, za investiciono odrzavanje reaktorskih sistema i nabavku opreme, za rekonstrukciju reaktorskih sistema. Ovaj izvestaj sadrzi 8 priloga koji opisuju rad reaktora, rad strucnih sluzbi i finansiranje.

  9. 75 FR 3761 - Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant, Units 1 and 2...

    Science.gov (United States)

    2010-01-22

    ... effluents that affect radiation exposures to plant workers and members of the public. Therefore, no changes... socioeconomic resources. Therefore, no changes to or different types of non-radiological environmental impacts...

  10. 75 FR 69137 - Southern Nuclear Operating Company Inc. Edwin I. Hatch Nuclear Plant, Unit No. 2 Environmental...

    Science.gov (United States)

    2010-11-10

    ... no change to radioactive effluents that affect radiation exposures to plant workers and members of... socioeconomic conditions in the region. Therefore, no changes to or different types of non-radiological...

  11. Argentinian Experience and Perspectives for Small and Medium Sized Reactors. Annex I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

    The National Atomic Energy Commission (CNEA) was created on 31 May 1950. It was responsible for all nuclear activities in the country until 1994. In 1994, the regulatory body and the operational activities of nuclear power plants were separated. A utility named Nucleoelectrica Argentina (NASA) was created. The present missions of the CNEA are: - To assist the national Government in nuclear policy; - To provide research and development in nuclear areas, including nuclear power plants, research reactors and nuclear fuel cycles; - To provide spent fuel and radioactive waste management, decommissioning of nuclear and radioactive installations, and environmental remediation; - By itself or through related companies, to provide radioisotopes for medicine and industry, and provide services to nuclear power plants and conventional industries. Development of human resources is yet another important objective of CNEA activities.

  12. Spectrographic Determination of Impurities in Ceramic Materials for Nuclear Fusion Reactors. I. Analysis of Alumina

    International Nuclear Information System (INIS)

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

    1990-01-01

    The determination of minor and trace elements in the aluminium oxide considered as possible ceramic material in thermonuclear fusion reactors has been studied. The concentration ranges are 0.1 - 0.3 * for Ca, Si and Y, and at the ppm level for Co, Cr, Fe, Hf, K, Li, Mg, Mn, Na, Ni, Se, Ta, Ti, V and Zr. Atomic emission spectroscopy with direct current ore excitation and photographic detection has been employed. For Hf, Mg, Ta, Ti, V and Zr the use of 40% of copper fluoride as a carrier and of Nb as lnternal standard provide suitable sensitivities and precessions, while for the rest of elements the bent results are obtained with graphite powder in different proportions and Rb or Sn as internal standard. (Author) 7 refs

  13. Neutron radiography applications in I.T.U. TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, A. B.

    2002-01-01

    Neutron radiography is an important radiographic technique which is supplied different and advanced information according to the X or gamma ray radiography. However, it has a trouble for supplying the convenient neutron sources. Tangential beam tube of Istanbul Technical University (ITU) TRIGA Mark-II Training and Research Reactor has been arranged for using neutron radiography. The neutron radiography set defined as detailed for the application of the technique. Two different techniques for neutron radiography are defined as namely, transfer method and direct method. For the transfer method dysprosium and indium screens are used in the study. But, dysprosium generally was preferred in many studies in the point of view nuclear safety. Gadolinium was used for direct method. Two techniques are compared and explained the preferring of the transfer technique. Firstly, reference composition is prepared for seeing the differences between neutron and X-ray or gamma radiography. In addition of it, some radiograph samples are given neutron and X-ray radiography which shows the different image characters. Lastly, some examples are given from archaeometric studies. One of them the brass plates of Great Mosque door in Cizre. After the neutron radiography application, organic dye traces are noticed. Other study is on a sword that belong to Urartu period at the first millennium B.C. It is seen that some wooden part on it. Some different artefacts are examined with neutron radiography from the Ikiztepe excavation site, then some animal post parts are recognized on them. One of them is sword and sheath which are corroded together. After the neutron radiography application, it can be noticed that there are a cloth between the sword and its sheath. By using neutron radiography, many interesting and detailed results are observed in ITU TRIGA Mark-II Training and Research Reactor. Some of them shouldn't be recognised by using any other technique

  14. Recently-developed neutron activation analysis techniques utilizing the University of California at Irvine TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Guinn, V.P.; Chambless, D.; Cortes T, E.; DeLancey, K.; Garzonov, E.; Miller, D.A.; Miller, G.E.; Purcell, M.A.

    1976-01-01

    The University of California at Irvine (UCI) 250 kW TRIGA Mark I reactor is used extensively for neutron activation analysis (NAA) studies. These particularly include basic technique studies and application studies in the fields of environmental pollution, crime investigation, archaeology, oceanography, and geochemistry. In recent NAA studies at UCI, a number of techniques have been developed which considerably improve the usefulness of such a research reactor for NAA work, and which should be of interest and use to others. Six of these techniques will be described in further detail in the full paper. They are as follows: development and use of (1) an automated high-precision rapid transfer system for instrumental NAA measurements with induced activities having half lives as short as 0.5 second, (2) an automated measurement system and computer program for making accurate dead-time corrections under conditions where the Ge(Li) spectrometer deadtime is changing rapidly during the counting period, (3) a technique to minimize the loss of mercury from samples during reactor irradiation via the use of dry-ice-packed, vented TRIGA rotary rack tubes, (4) a technique for compacting powdered samples, by pre-irradiation treatment with a solution of paraffin in carbon disulfide, to provide reproducible irradiation and counting geometries, (5) a method utilizing hydrated antimony pentoxide (HAP) as a pre-irradiation treatment material for removal of sodium from aqueous and wet-ashed samples, and (6) a computerized system for predicting in advance of activation, from approximate known elemental compositions, the total counting rate, deadtime, spectrum shape, principal photopeaks, and approximate actual lower limits of instrumental NAA detection of designated elements for any selected irradiation and decay times. (author)

  15. Possible divertor solutions for a fusion reactor. Pt. I. Physical aspects based on present day divertor operation

    International Nuclear Information System (INIS)

    Kallenbach, A.; Bosch, H.-S.; De Pena Hempel, S.; Dux, R.; Kaufmann, M.; Mertens, V.; Neuhauser, J.; Suttrop, W.; Zohm, H.

    1997-01-01

    For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m -2 , far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Z eff . Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δP rad /δZ eff ∝n e 2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)

  16. Electrochemical aspects on corrosion in Swedish reactor containments; Elektrokemiska aspekter paa korrosion i svenska reaktorinneslutningar

    Energy Technology Data Exchange (ETDEWEB)

    Ullberg, Mats [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

    2006-10-15

    Post-stressed concrete is used in all Swedish nuclear reactor containments. Steel in concrete is normally protected from corrosion by the highly alkaline pore solution in concrete. A passive film develops on the surface of steel in contact with the pore solution. However, corrosion may still occur under special circumstances. It is therefore desirable to monitor the corrosion status of the containment. A review of the corrosion experience with steel in concrete strongly suggests that the potential problem of most concern for the Swedish reactor containments is cavity formation during grouting of tendons and of penetrations in the containment wall. Cavities break the contact between alkaline grout and steel. Corrosion is then possible, provided the relative humidity is high enough. Normal methods for inspection of the corrosion status of steel reinforcement in concrete are not applicable to very heavy structures like reactor containments. Since inspections are difficult to carry out, it is important that they be focused on the most susceptible portions of the containment. This report is an attempt to assemble potentially useful background information. The original intention was to focus on electrochemical methods of investigation. When it was realized that the potential use of electrochemical methods was limited, the scope of the review was broadened. The present as well as previous investigations indicate that nondestructive testing of grouted tendons is the outstanding problem in the condition assessment of Swedish nuclear reactor containments. Grouted tendons are also used in a very large number of bridges built since the early 1950s. The experience gained in connection with bridges has therefore been investigated. The need for a testing method for grouted tendons in bridges has long been strongly felt and development work has been in progress since the early 1970-ies, for example within the Strategic Highway Research Project in the Unite States. Potential

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. Hardware resilience: a way to achieve reliability and safety in new nuclear reactors I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Divisão de Engenharia Nuclear. Serviço de Instrumentação; Nedjah, Nadia, E-mail: nadia@eng.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia de Sistemas e Telecomunicações

    2017-07-01

    The idea that systems have a property called ‘resilience’ has emerged in the last decade [1]. In this paper we intend to bring the idea of resilient systems for the hardware applied in safety-critical systems, such as the new nuclear reactor instrumentation and control (I and C) systems. The new systems (based in hardware description language (HDL) programmable devices) have been developed in response to the obsolescence of old analog technologies and current microprocessor-based digital technologies. Although HDL programmable devices have been widely used in various other industries for decades, they are still very new in nuclear reactors systems, which can be seen as a challenge and risk in the safety analyses and licensing efforts for utilities and designers. The goal of this work is to develop and test hardware architectures to tolerate the occurrence of faults, including multiple faults, minimizing the impact of the recovery process on system availability. Basic concepts of resilience in complex systems, as 'return to equilibrium', 'robustness' and 'extra adaptive capacity' were analyzed from the point of view of hardware architectures, leading to linkages between concepts and methods for resilience using an approach that increases reliability and simplifies the licensing process of systems based in HDL programmable devices in nuclear plants. (author)

  19. Hardware resilience: a way to achieve reliability and safety in new nuclear reactors I and C systems

    International Nuclear Information System (INIS)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Nedjah, Nadia

    2017-01-01

    The idea that systems have a property called ‘resilience’ has emerged in the last decade [1]. In this paper we intend to bring the idea of resilient systems for the hardware applied in safety-critical systems, such as the new nuclear reactor instrumentation and control (I and C) systems. The new systems (based in hardware description language (HDL) programmable devices) have been developed in response to the obsolescence of old analog technologies and current microprocessor-based digital technologies. Although HDL programmable devices have been widely used in various other industries for decades, they are still very new in nuclear reactors systems, which can be seen as a challenge and risk in the safety analyses and licensing efforts for utilities and designers. The goal of this work is to develop and test hardware architectures to tolerate the occurrence of faults, including multiple faults, minimizing the impact of the recovery process on system availability. Basic concepts of resilience in complex systems, as 'return to equilibrium', 'robustness' and 'extra adaptive capacity' were analyzed from the point of view of hardware architectures, leading to linkages between concepts and methods for resilience using an approach that increases reliability and simplifies the licensing process of systems based in HDL programmable devices in nuclear plants. (author)

  20. Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

    International Nuclear Information System (INIS)

    Saavedra, Fernando M.

    2001-01-01

    The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

  1. RA Reactor operation and maintenance (I-IX), part VIII, Task 3.08/05, Decontamination of the reactor; Pogon i odrzavanje reaktora (I-IX), VIII Deo, Zadatak 3.08/05 Dekontaminacija reaktora

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    Permanent increase of radiation in the heavy water system was noticed during first three year of the RA reactor operation, even when the reactor was shutdown. It was found that there was no failure of the fuel element cladding. Radioactive cobalt was found in the heavy water which was rather strange. During repair of the heavy water system, it has been found that stellite was used for coating the heavy water pumps. Since stellite is a cobalt alloy, this could have been the source of radioactive cobalt in the heavy water. The stellite coating was damaged due to friction and particle of cobalt appeared in the coolant, they were activated since they were in the core. decontamination of the heavy water and the heavy water coolant loop was a must . Beside the detailed report on the contamination and decontamination of the heavy water system this volume includes 14 annexes describing the investigation of the event and the whole procedure of decontamination.

  2. Determination of the effective delayed neutron fraction in the Coral-I Reactor; Determinacion de la fraccion efectiva de neutrones retardados en el Reactor Coral-1

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E

    1973-07-01

    The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

  3. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters; Fizicka merenja na reaktoru RA u vezi projekta VISA-2 - I deo, Pustanje u rad reaktora RA i merenje fizickih parametara novog jezgra reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included. [Serbo-Croat] Svrha merenja je odredjivanje neutronskog fluksa u reaktoru RA. S obzirom na uvecani broj tehnoloskih kanala of 56 na 68 u vezi projekta VISA-2, bilo je potrebno ponovo dovesti reaktora RA do kriticnosti i izvrsiti merenja karakteristika fluksa neutrona. Posebno je pripremljen 'program pustanja u pogon reaktora RA', koji je sadrzan u ovom dokumentu. Program merenja bio je podeljen na dve faze. Prva faza je merenje fluksa pre podizanju reaktora na nominalnu snagu. Slicna merenja vrsena su i na vecim snagama u drugoj fazi, pod uslovima ravnoteznog zatrovanja reaktora ksenonom, jer se tada pokazuju izvesne promene u odgovarajucim karakteristikama fluksa neutrona. Ovaj izvestaj sadrzi merene vrednosti raspodele fluksa i apsolutne vrednosti termalnih i brzih neutrona kao i kadmijumskih odnosa koji su korisceni za odredjivanje fluksa epitermalnih neutrona. Opisana je kalibracija regulacionih sipki za hladan nezatrovan reaktor.

  4. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    biggest difficulty was maintenance of reactor instrumentation. During 1981 the reactor was operated safely, there was no accident nor incident that would affect the safety of reactor personnel or the environment. The testing operation will be continued in 1982,and the experience so far shows that the program would be successfully fulfilled on the whole. [Serbo-Croat] Nuklearni reaktor RA prestao je sa radom nakon martovske kampanje 1979. godine usled pojave talozenja oksihidrata aluminijuma na kosuljicama gorivnih elemenata. Odgovarajucim resenjima Sanitarnog inspektorata Republickog sekretarijata za zdravje i socijalnu politiku SR Srbije i generalnog direktora Instituta za nuklearne nauke 'Boris Kidric', Vinca zabranjen je dalji rad reaktora sve dok se ne utvrde uzroci stvaranja oksihidrata aluminijuma i njihovog talozenja, preduzmu mere za njihovo uklanjanje i ne obezbede potrebni uslovi za normalan nastavak rada reaktora. Do kraja 1979. i tokom 1980. godine, nakon niza izvrsenih analiza i utvrdjivanja uzroka koji su doveli do zaustavljanja rada reaktora, izvrsene su sve neophodne pripreme za ponovno pustanje reaktora u rad. Polazeci od cinjenice da na reaktoru RA ne postoji sistem za hladjenje jezgra u slucaju udesa i da ne postoji adekvatan sistem za filtriranje potencijalno zagadjenog vazduha, a saglasno sa novim propisima o pustanju u rad i probnom radu nuklearnih objekata, Sanitarni inspektorat je doneo privremeno resenje kojim se dozvoljava pustanje reaktora u rad, tj. izvodjenje tzv. 'nultog eksperimenta' uz ogranicenje snage na 1% od vrednosti nominalne snage. Na osnovu dobijene dozvole, reaktor RA je ponovo pusten u rad 21. januara 1981. godine, kada je dostignuta kriticnost sa jezgrom sastavljenim iskljucivo od gorivnih elemenata od 80% obogacenog uranijuma. Eksperiment je zavrsen krajem marta, nakon cega je zatrazena dozvola za probni rad na vecim snagama i potom za rad na punoj snazi. Uzimajuci postojece stanje reaktora RA doneto je resenje kojim se

  5. Research nuclear reactor RA - Annual Report 1975. Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1975. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    The plan for 1975 was successfully fulfilled. This is reflected in research related to improvement of operating properties of the RA reactor, mostly due to the effort of the RA staff employed in operation and maintenance of the reactor. Fuel saving achieved by this activity amounted to about 38% (80% enriched fuel). Preliminary work is done, concerned with new reactor core with highly enriched fuel. This is a significant saving as well. New fuel elements have arrived at the end of this year. It is going to enable increase of neutron flux by 50% without changing the nominal operating power. The possibility of further improvement of the reactor are analyzed, to enable material testing and production of radioactive sources. Mid term plan for reactor operation was made according to this analysis. It is planned to further increase the neutron flux in isolated smaller zones, and building new experimental loops with cooling and fast neutron converters. Much was done to increase the safety level of reactor operation and preparing the safety report. [Serbo-Croat] Izvrsenje zadataka u 1975. godini bilo je uspesno. To se ogleda u povecanju istrazivackog rada vezanog za poboljsanje eksploatacionih karakteristika reaktora RA, pretezno koriscenjem sopstvenog kadra angazovanog u pogonu i odrzavanju reaktora. Ovim radom postignuta je usteda goriva od oko 38% (80% obogaceno gorivo). Izvrseni su preliminarni radovi na prevodjenju reaktora RA na novo gorivo, sto je takodje velika usteda. Novo gorivo je stiglo krajem godine i ono ce obezbediti porast neutronskog fluksa od 50%, bez promene nominalne snage reaktora. Izvrsena je analiza mogucnosti daljeg usavrsavanja reaktora za potrebe ispitivanja materijala kao i proizvodnju radioaktivnih izvora. Na osnovu ove analize nacinjen je srednjorocni program rada reaktora RA sa tezistem na daljem povecanju fluksa u izdvojenim manjim zonama i ugradnju 'hladjenih petlji' i brzih konvertora. Mnogo je ucinjeno na povecanju stepena sigurnosti

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-01-15

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

  7. Google tegemas taas võimsat kvartalit / Fredy-Edwin Esse

    Index Scriptorium Estoniae

    Esse, Fredy-Edwin

    2011-01-01

    Nii Google'i kasum kui ka käive on prognooside kohaselt järsult tõusnud, prognoositakse 6,3 mld. dollari suuruse käibe juures kasumiks 8,14 dollarit aktsia kohta. Autor panustaks oma raha Google'isse vaid lühiajaliselt

  8. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1988. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    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. [Serbo-Croat] Prema planu za 1988. godinu, reaktor RA je trebalo da pusten u rad oktobra meseca, medjutim nije dobio dozvolu za nastavak rada. Kontrola i odrzavanje opreme izvrsavani su redovno i efikasno, u granicama koje su diktirane raspolozivoscu repromaterijala i rezervnih delova. Najvecu poteskocu pricinjavalo je odrzavanje instrumentacije. Period stajanja u 1988. godini iskoriscen je za remont teskovodnih pumpi u primarnom kolu hladjenja. U cilju povecanja pouzdanosti rada reaktora zapoceti su radovi na modernizaciji instrumentacije, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, sklopljen je ugovor o izradi ove opreme koja bi trebalo da bude isporucena do kraja 1990. U cilju povecanja prostora za skladistenje ozracenog

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  10. Research reactor core conversion guidebook. V. 4: Fuels (Appendices I-K)

    International Nuclear Information System (INIS)

    1992-04-01

    Volume 4 consists of detailed Appendices I-K, which contain useful information on the properties, irradiation testing, and specifications and inspection procedures for fuels with reduced uranium enrichments. Summaries of these appendices can be found in Chapters 9-11 of Volume 1 of this guidebook. Refs, figs, tabs and samples

  11. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

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

  12. Reactor container

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  13. RA Research reactor, Annual report 1973; Istrazivacki nuklearni reaktor RA - Izvestaj za 1973. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1973-12-15

    During 1973, RA reactor was operated at nominal power for 4687 hours and 54 hours at lower power levels. The total production was 30504 MWh which is 1.6% higher than planned. Practically there was no discrepancies from the plan, since the action plan was corrected at the beginning of this year caused by the demand of changing the schedule for refuelling for the purpose of 'power excursion' experiment. The reactor was used for irradiation and experiments according to the demand of 336 users. This report contains detailed data about reactor power and experiments performed in 1973. Total number safety shutdowns was 12, of which 7 were caused by power cuts. Three shutdowns caused by failures of the equipment were caused by failures of new electronic tubes. Two shutdowns were caused by the operators. There have been three shorter interruptions announced power cuts. Total personnel exposure dose was lower than during previous years. There were no accidents during this year. Decontamination of surfaces was less than during previous years. Practically there was no surface contamination, and the quantity of collected radioactive waste was lower than previously. There were no liquid radioactive effluents. It was concluded that the successful operation in 1973 has a special significance taking into account the financial crisis. There still remains a number of unsolved problems related to: completing the inventory of spare parts, exchange of some elements of the equipment, exchange of instrumentation, and purchase of the highly enriched fuel. [Serbo-Croat] Reaktor RA je u 1973. godini radio na nominalnoj snazi 4687 sati i 54 sata na manjim snagama. Ukupni rad iznosio je 30504 MWh odnosno 1,6% vise od planiranog. Prakticno nije bilo odstupanja od plana rada koji je pocetkom godine korigovan zbog promene planiranih izmena goriva usled izvodjenja eksperimenta 'ekskurzije snage'. Reaktor je koriscen za ozracivanja i eksperimente za 336 korisnika. Ovaj izvestaj sadrzi detaljne

  14. PSEPLOT: a controller for plotting data from the Mark I Boiling Water Reactor Pressure Suppression Experiment

    International Nuclear Information System (INIS)

    Holman, G.S.

    1978-01-01

    PSEPLOT is a computer routine that was developed for the Lawrence Livermore Laboratory Octopus computer system to generate several thousand plots of engineering data in a consistent format for referencing and comparison. The time-dependent engineering data were recorded during each of 25 tests of the Mark I Pressure Suppression Experiment (PSE). Although PSEPLOT is restricted to PSE, its concept is applicable to any similar data management task

  15. The TEX-I real-time expert system applied to situation assessment for the SNR-300 reactor

    International Nuclear Information System (INIS)

    Schmal, N.; Doerbecker, K.; Leder, H.J.; Rueckert, M.; Schade, H.J.

    1990-01-01

    Within the German TEX-I Project, which was sponsored by the Federal Ministry for Research and Technology, several companies developed industrial applications of technical expert systems for data interpretation, diagnosis and process control. The purpose of the diagnosis expert system reported here is to support the operators of the LMFBR SNR-300 in assessing plant status in real-time, based on readings from a large number of sensors. By online connection to the process control computer, it can monitor all incoming signal values, check the consistency of data, continuously diagnose the current plant status, detect unusual trends prior to accidents, localize faulty components and recommend operators response in abnormal conditions. The systems architecture consists of two basic subsystems, an inference engine and an intelligent process interface, implemented in Lisp on a Symbolics-Workstation. The inference engine has been derived from BABYLON, a hybrid shell developed by the German computer research institute GMD. This shell includes rules, prologue, constraints and an object oriented frame processor. The extended version has a component description language and a top-down diagnosis scheme including a mechanism of attention focusing. Inference run as independent, quasi-parallel processes. These so-called inference tasks can interrupt or abort each other, if higher priority events must be processed. The complete system is modelled in an object oriented matter and is divided into several subsystems and each subsystem into its physical components. At present the expert system is connected to a real-time simulation of the reactor. The simulation is based on a thermohydraulic code for simulation of the transient behaviour of temperatures and flow rates in the reactor core, plena, pipes, pumps, valves, intermediate heat exchangers and cooling components. Additionally, the systems response to an asynchronous operator interaction can be simulated. Several types of anomalies

  16. Final air test results for the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Collins, E.K.; Lai, W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water reactor (BWR) power plant has never occurred. However, because this type of accident is particularly severe, it is used as a principal basis for design. During a hypothetical LOCA in a Mark I BWR, air followed by steam is injected from a drywell into a toroidal wetwell about half-filled with water. A series of consistent, versatile, and accurate air-water tests simulating LOCA conditions was completed in the Lawrence Livermore Laboratory 1/5-Scale Mark I BWR Pressure Suppression Experimental Facility. Results from this test series were used to quantify the vertical loading function and to study the associated fluid dynamic phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variations of hydrodynamic-generated vertical loads with changes in drywell pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1/5-scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings which are invariant. These groupongs show that if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor; the time when these forces occur is reduced by the square root of the scale factor

  17. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  18. Insertion of control systems models in the Almod 3 computer code for the simulation of Angra I reactor start-up tests

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1981-09-01

    The Almod 3 computer code was modified, aiming at the simulation of Angra I nuclear power plant behavior during some reactor start-up tests. The results obtained with the modified computer code (Almod 3W) are compared with those obtained with the Retran computer code. (E.G.) [pt

  19. Measurement of graphite and aluminium absorption cross sections via reactor period by danger coefficient method; Merenje apsorpcionih preseka grafita i aluminijuma preko periode reaktora metodom koeficijenta opasnosti

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Markovic, V; Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1963-07-01

    Full text: This activity is a logical continuation of the experiment at the RA reactor during 1962 which was based on compensating the effect by means of control rod. Since results are given with significant errors, new method for measuring the absorption cross sections via reactor period. Experiment was done at the RB reactor which was particularly prepared for this type of experiments. Reactor power was from 50 mW to 2 W. Absorption cross sections were measured for two types of material: domestic graphite No.3 and French graphite 'Pachiney', and two types of aluminium. Total errors in applying this method are {+-} 5%, where the source of major part of error comes from uncertainty of the standard absorption power (previous method gave {+-} 10 do 55% ). Comparison of French graphite absorption cross section obtained via reactor period and via control rod showed approximate agreement with discrepancy of 5.4% which is considered within the precision of this method. Considering the accuracy of measurement results and reactor economy it is concluded that measuring absorption cross sections of samples via period of RB reactor is more favourable than measurements by control rod at the RA reactor. Pun tekst: Ovaj rad predstavlja logican nastavak eksperimenta na reaktoru RA u toku 1962. godine, koji je bazirao na kompenzaciji efekta pomocu kontrolne sipke. Kako su rezultati dati sa velikim greskama, to se prislo novom nacinu merenja apsorpsionih preseka preko periode reaktora. Eksperiment je radjen na reaktoru RB koji je specijalno pripremljen za ovu vrstu eksperimenta. Snaga reaktora se kretala od 50 mW do 2 W. Preko periode reaktora RB odredjeni su apsorpcioni preseci za dve vrste materijala i to: domaci grafit No.3 i francuski 'Pachiney', i dve vrste aluminijuma. Ukupne greske pri ovom nacimu merenja iznose oko {+-} 5%, gde glavni deo greske nosi neodredjenost apsorpcione moci standarda (ranija metoda je dala {+-} 10 do 55% ). Poredjenjem vrednosti apsorpcionih preseka

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

    International Nuclear Information System (INIS)

    Diaz, A.Z.

    2001-01-01

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

  1. Photocatalytic Membrane Reactor for the Removal of C.I. Disperse Red 73

    Directory of Open Access Journals (Sweden)

    Valentina Buscio

    2015-06-01

    Full Text Available After the dyeing process, part of the dyes used to color textile materials are not fixed into the substrate and are discharged into wastewater as residual dyes. In this study, a heterogeneous photocatalytic process combined with microfiltration has been investigated for the removal of C.I. Disperse Red 73 from synthetic textile effluents. The titanium dioxide (TiO2 Aeroxide P25 was selected as photocatalyst. The photocatalytic treatment achieved between 60% and 90% of dye degradation and up to 98% chemical oxygen demand (COD removal. The influence of different parameters on photocatalytic degradation was studied: pH, initial photocatalyst loading, and dye concentration. The best conditions for dye degradation were pH 4, an initial dye concentration of 50 mg·L−1, and a TiO2 loading of 2 g·L−1. The photocatalytic membrane treatment provided a high quality permeate, which can be reused.

  2. Operation, maintenance and utilization of the RA reactor - Report on operation in 1979; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1979. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M

    1979-12-15

    During 1979 the RA reactor was in operation only three months, i.e. only 24% of the planned activity was achieved. The reactor operation was interrupted in March due to problems that could not be solved by the existing equipment. Alkalinity of the heavy water resulting from the existing ammonia ions could not be removed by the existing distillation system. In addition deposition of aluminium oxyhydrate on the fuel elements was increased. This was noticed during routine control when the reactor was shutdown for refueling. The decision of the Director general of the Institute and sanitary Inspector followed, prohibiting further reactor operation. A separate chapter of this report is devoted to the analysis of the difficulties and possible solution of the problem in cooperation with the experts from different laboratories of the Institute. Aged and damaged instruments at the reactor were not exchange due to lack of budget. During the operation there were no accidents. [Serbo-Croat] U toku 1979. godine plan rada ostvaren je sa 24%, odnosno reaktor je radio samo prva tri meseca. U martu je doslo do prekida rada reaktora zbog pojave koja se nije mogla otkloniti postojecom opremom. Alkalnost teske vode, posledica prisustva amonijum jona, ne moze se odstraniti sistemom destilacije. Pored toga usled dotrajalosti opreme povecano je talozenje aluminijum oksihidroksida na gorivnim elementima. Ova pojava uocena je tokom rutinske kontrole prilikom izmene goriva a usledila je obustava rada resenjem Direktora i sanitarnog inspektora. Posebno poglavlje ovog izvestaja posveceno je analizi teskoca u radu i reaktora i resavanju nastalih problema sa gorivom u saradnji sa saradnicima drugih laboratorija Instituta. Dotrajali uredjaji i oprema nisu zamenjeni usled nedostatka sredstava. U toku rada nije bilo akcidenata.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-12-15

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

  4. Gentzler, Edwin. Translation, hypertext, and creativity: Contemporary translation theories. Bristol: Multilingual Matters, 2001. 232 p.

    Directory of Open Access Journals (Sweden)

    Davi S. Gonçalves

    2018-01-01

    Full Text Available Contemporary translation theories (Gentzler,2001 provides readers with a thorough historical analysis of how the notion of creativity and autonomy in what regards reading has been transformed – as well as regarding its influence towards the idea of translation. The place occupied by the translator is a place between spaces; a fluid locale where any concreteness has melted. Meaning is thus not graspable or amenable to be tamed; on the contrary, literature is about opening up more space for the wilderness to be (rediscovered. A text is many texts, a hypertext, filled in with narratives that mutually supplement one another, deconstructing and reconstructing meanings; and, within such picture, translation emerges not as an opportunity to resurrect the body of an original text, but as a phantasm of both sameness and uniqueness. What does exist cannot be seen; it is always on the run; meanings surface from liquefied pages, pages that escape our attempt of defining them for good. This is why translation can be taken as metonym: as s/he recreates the original text within the target context, the translator choose to highlight those textual elements that s/he deems relevant – those fragments of the text that have touched and determined his/her reading. The experience of translation, that goes beyond dichotomist standards (e.g. foreign/domestic, equivalent/adapted, etc., is finally taken as a profitable realm for the literary discourse to validate its impalpability. Such shift in the approach towards translation is significant because, even though the process of recreation takes place in every textual practice, tradition has been pressuring translation scholars towards the designing of guidelines and norms that, I dare say, only obstruct the task of translating.

  5. New neutron and gamma dosimetry equipment at the RB reactor; Nova merna neutronska i gama dozimetrijska oprema na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-15

    In the frame of bilateral cooperation between Germany and Yugoslavia, complete control, safety and dosimetry equipment of the shut-down SNEAK reactor was donated to Vinca Institute and transported to be installed at the RB reactor. This report contains detailed description of instrumentation components including detectors, electronic components and electronic circuits. Experimental data which verified correct functioning of the installed devices are part of this document. The objective of the RB reactor staff is to achieve new safety and dosimetry system in order to improve the reliability and availability of the RB reactor for future experiments.

  6. Source term attenuation by water in the Mark I boiling water reactor drywell

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States)

    1993-09-01

    Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000.

  7. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  8. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

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

  9. Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac No. 8532, Col. Progreso, C.P. 62550, Jiutepec, Morelos (Mexico)

    2012-07-01

    The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention was focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)

  10. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  11. Operation, maintenance and utilization of the RA reactor, Annual report 1978; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1978. godini

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-15

    It has been planned for 1978 that the RA reactor would be operated for 158 dana at nominal power of 6.5 MW meaning production of 24 648 MWh. The plan was fulfilled since 24 652 MWh was produces. Reactor operation for 158 days is relevant to reactor operation for 200 days in the period before 1975. The reason is increased neutron flux achieved due to improved fuel management and the characteristics of the new 80% enriched fuel. At the end of 1978 the reactor core contained 45% of 80% enriched fuel elements. Increase of neutron flux has shortened the typical time needed for irradiation of the most important samples for isotope production. This significant success in reactor operation is at the same time an obligation for increasing its utilization. Some new trends proposed for increasing reactor utilization capacities were presented at the Conference on utilization of research nuclear reactors in Yugoslavia held in May 1978. [Serbo-Croat] Reaktor RA imao je u planu za 1978. godinu 158 dana rada na nominalnoj snazi od 6.5 MW, sto odgovara radu od 24 648 MWh. Ostvareno je 24 652 MWh sto znaci da je plan ostvaren. Rad reaktora od 158 dana odgovara radu reaktora od 200 dana u periodu pre 1975. godine. Razlog je povecanje neutronskog fluksa zahvaljujuci usavrsenom rukovanju gorivom i karakteristikama novog 80% obogacenog goriva. Krajem 1978. godine 45% jezgra reaktora bilo je popunjeno novim 80% obogacenim gorivom. Povecani neutronski fluks omogucio je skracenje vremena ozracivanja vaznih uzoraka za proizvodnju radioaktivnih izotopa. Ovaj znacajan uspeh je istovremeno obaveza znatno veceg iskoriscenja reaktora RA. Rezultati napora da se postigne vece iskoriscenje reaktora RA prezentirani su na Konferenciji o koriscenju nuklearnih reaktora u Jugoslaviji koja je odrzana u maju 1978.

  12. Radiation protection at the RA Reactor in 1995, Part -2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste; Deo 2 - Prilog 2 - Dekontaminacija i intervencije, skupljanje tecnih efluenata i cvrstih radioaktivnih otpadnih materijala

    Energy Technology Data Exchange (ETDEWEB)

    Mandic, M; Vukovic, Z; Lazic, S; Plecas, I; Voko, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1995-12-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [Serbo-Croat] Tokom rada reaktora RA dolazi do stvaranja odredjenih cvrstih otpadnih materijala cija prosecna kolicina zavisi od vremena rada reaktora i aktivnosti koje se tamo obavljaju. Tokom remonta, kada reaktor ne radi kao i pri akcidentalnim situacijama nastaju vece kolicine otpadnih materijala koje zavise od obima i vrste remontnih operacija i obima dekontaminacije kontaminirane radne povrsine i kontaminiranog alata, predmeta, opreme, itd. Nastali otpadni materijali se razvrstavaju i pakuju na mestu nastanka prema odgovarajucim propisima u skladu sa principima zastite od zracenja i aspekta bezbednosti u cilju minimiziranja nepotrebnog ozracivanja ljudstva za preuzimanje, kontrolu, transport, naknadnu obradu RAO i dekontaminaciju. Pri nerutinskim operacijama (dekontaminacija, remont, kontaminiarni otpadni materijal velike zapremine i sl.), strucna sluzba Institita ZASTITA pruza strucne konsultacije i pomaze pri planiranju

  13. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    International Nuclear Information System (INIS)

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-10-15

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

  15. Operation, maintenance and utilization of the RA reactor, Annual report for 1980; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1980. godini

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-12-15

    During 1980 the activities of RA reactor staff was conducted in two directions: repair works and determination of the state of the existing equipment; and preparing and constructing additional equipment and completing the documentation needed to fulfill the new legal regulations. preparation of the existing equipment for future operation was finished, but construction of the additional equipment (according to the regulations) is not done at planned rate due to different reasons such as lack of budget, administration, import. This report includes a chapter devoted to the analysis of the operation difficulties that caused prohibition of reactor operation as well as difficulties in achievement of the planned activities for 1980. Data about financial issues are included as well. [Serbo-Croat] U toku 1980. godine rad OOUR-a Nuklearni reaktor RA odvijao se u dva smera: remontnim radovima i utvrdjivanju stanja postojece opreme i u pripremi i ugradnji potrebne dodatne opreme i kompletiranju neophodne dokumentacije prema novim zakonskim propisima. Priprema postojece opreme za rad je u potpunosti izvrsena, dok se projekti i ugradnja dodatne opreme (prema zakonskim obavezama) ne odvija potrebnom brzinom zbog raznih objektivnih razloga, kao nedostatak sredstava, odobrenja, uvoza. Ovaj izvestaj sadrzi poglavlje posveceno analizi teskoca u radu koje su dovele do zabrane rada reaktora, i teskoca u realizovanju plana rada u 1980. godini, ako i podatke o finansiranju projekta 'pogon, odrzavanje i eksploatacija reaktora RA'.

  16. Edwin Powell Hubble

    Indian Academy of Sciences (India)

    Srimath

    lum inosity relationship for C epheids,he proved that these nebulae w ere far aw ay,and had to be galaxies that lie beyond the M ilky W ay. H is 疸dings w ere revolutionary and changed the existing view of the U niverse,w hen it w as thought that allthe fuzzy nebulae w ere part ofthe M ilky W ay. H ubble also devised a classi ...

  17. Edwin – Wosu, N

    African Journals Online (AJOL)

    USER21

    SPECIES OF CUCURBITS AMONG THE PEOPLE OF NIGER DELTA. NIGERIA ... –sized prostrate runner, primarily found in the Warmer regions of the world, and a ... bounding the western periphery and on the south by the. Atlantic Ocean .... various part of the world, as has been noted in Asian continent and some West ...

  18. Edwin Powell Hubble

    Indian Academy of Sciences (India)

    Srimath

    ... and hence the U niverse m ust expand or contract depending on the available m atter. B ut since the idea app eared so far-fetched,E instein m odi¯ed his equations so that the U niverse rem ains static. E dw in H ubble's observations revolutionized astronom y,pre- senting the ¯rst evidence that the U niverse is expanding.

  19. Actions needed for RA reactor exploitation - I-IV, Part I, Thermotechnical experiments related to RA reactor hot start-up; Radovi za potrebe eksploatacije reaktora RA - I-IV, I Deo, Termotehnicki eksperimenti u vezi pustanja rektora RA na snagu

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-12-15

    Heavy water coolant loop of the RA reactor includes the reactor, circulation pumps, heat exchangers and pipes. The objective of this task was measuring the thermal parameters of the RA reactor during operation. This report contains the results of the experiment, calculations of thermal regime for the outer and inner tubes, maximum temperature of the fuel element, fluid flow rate in the reactor channels, temperature of the coolant and fuel element cladding.

  20. Safety shutdowns and failures of the RA reactor equipment; Sigurnosna zaustavljanja i kvarovi opreme na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Mitrovic, S [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

    1966-07-01

    This report is an attempt of statistical analysis of the failures occurred during RA reactor operation. A list of failures occurred on the RA equipment during 1965 is included. Failures were related to the following systems: dosimetry system (22%), safety and control system (7%), heavy water system (2%), technical water (4%), helium system (2%), measuring instruments (30%), transport, ventilation, power supply systems (32%). A review of safety shutdowns from 1962 to 1966 is included as well, as a comparison with three similar reactors. Although the number of events used for statistical analysis was not adequate, it has been concluded that RA reactor operation was stable and reliable.

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

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-02-01

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

  2. Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Raisic, N; Obradovic, D; Jovanovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-05-01

    A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness.

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

    International Nuclear Information System (INIS)

    Mankamo, T.

    1996-04-01

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

  4. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

  5. Domestic wastewater anaerobic treatment I : Performance of one-step UASB and HUSB reactors; Tratamiento anaerobio de aguas residuales urbanas I : Aplicacion de reactores UASB y HUSB de etapa unica

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez Rodriguez, J. A.; Gomez Lopez, M.; Soto Castineira, M.

    2005-07-01

    Domestic wastewater treatment was carried out on a pilot scale anaerobic digester, with an active volume of 25.5 m''3. The digester operated at different conditions: (a) as an UASB reactor (up-flow anaerobic sludge blanket), with the aim of reaching a complete anaerobic treatment of domestic wastewater, and (b) as a HUSB (hydrolytic upflow sludge blanket) reactor, working in this case as a wastewater pre-treatment that removes suspended solid matter and increase the effluent biodegradability. The advantages of these treatment systems are its economic feasibility, no energy consumption and low excess sludge generation. (Author) 17 refs.

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  7. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  8. Data concerning operation and application of RA reactor in 1975, Annex 1; Prilog 1 - Podaci o radu i iskoriscenosti reaktora RA u 1975. godini

    Energy Technology Data Exchange (ETDEWEB)

    Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    RA reactor was operating according to the plan for 1975 adopted in December of the previous year. It was planned for reactor to be operated at nominal power first 10-20 days each month, three following days were reserved for different power levels according to the users' demand. Four fuel exchanges were planned and fulfilled with minor delay. Data concerning planned and real operation, as well as delays from the plan and shorter interruptions are presented in tables of this Annex. It is shown that all the delays and interruptions which amounted to 104 hours were compensated. [Serbo-Croat] Reaktor RA je radio prema planu rada za 1975. godinu, nacinjenom u decembru prethodne godine. Planirano je da reaktor radi neprekidno prvih 10-20 dana u mesecu na nominalnoj snazi, tri sledeca dana je rezervisano za rad na drugim snagama zavisno od potreba korisnuka. Planirane su i 4 izmene goriva. Podaci o planiranom i ostvarenom radu kao i odstupanjima od plana i kracim prekidima u radu reaktora dati su u tabelama ovog priloga. Vidi se da su sva odstupanja i prekidi u ukupnom trajanju od 104 sata u celini nadoknadjeni.

  9. An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► The paper identifies possible transient and accident scenarios in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Key accidents scenarios were investigated through qualitative reasoning. ► The accidents were found to constitute loss of heat sink event for the nuclear reactor. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Quantitative study of the possible operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven unique case studies are proposed based on a thorough review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. The qualitative parameters of each case study are outlined as well as the basis in literature. A previously published modeling scheme is described and adapted for application as a simulation platform for these transient events. The results of the quantitative case studies are described within part II of this paper.

  10. Description of the control and safety systems of the RA reactor; Opis sistema za upravljanje i sigurnosnu zastitu RA

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, B; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1962-07-01

    This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D{sub 2}O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally.

  11. Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, V Deo, Odredjivanje fine raspodele fluksa

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-15

    Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S{sub n} code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S{sub 4} approximation. Reactor cell was treated as D{sub 2}O-U-D{sub 2}O system. Time of iteration was 185 s. Resena je transportna monoenergetska jednacina numerickom metodom Carlsona u cilindricnoj geometriji. Razvijanje S{sub n} kod za digitalnu masinu ZUSE-Z 23. Odredjena je raspodela fluksa u celiji reaktora RA S{sub 4} aproksimacijom. Celija je tretirana kao D{sub 2}O-U-D{sub 2}O. Vreme iteracije je 185 sec (author)

  12. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, IV Deo, Promena izotopnog sastava goriva

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-07-15

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U{sup 235} chain, analytical expressions for the concentrations of U{sup 235}, U{sup 236} and Np{sup 237} as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer.

  13. Dosimetry and technical radiation protection at the RA Reactor - Report for 1977, Annex V; Prilog V - Dozimetrija i tehnicka zastita od zracenja kod reaktora RA - Izvestaj za 1977. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    This report includes data about the level of gamma and neutron radiation, and level of contamination in the working environment at the RA reactor, data about collective and individual exposure of the staff to radiation, data about contamination and decontamination as well as radioactive waste. During 1977, at the RA reactor there was no accident that would cause significant exposure of tf the staff or contamination of the working space and the environment of the reactor. [Serbo-Croat] Ovaj Izvestaj sadrzi podatke o nivou gama i neutronskog zracenja i stepenu kontaminacije radne sredine, podatke o individualnom i kolektivnom izlaganju zracenju radnog osoblja, podatke o kontaminaciji i dekontaminaciji kao i o radioaktivnom otpadu. U toku 1977. godine, na reaktoru RA nije bilo akcidenata vecih razmera koji bi za posledice imali znacajnije ozracivanje radnog osoblja i kontaminaciju radne sredine i okoline.

  14. Irradiation of reactor materials within projects VISA-2 and 3, 3. Procedure for construction and testing the capsules and test-tubes - Phase I (Parts I and II) Part II; Ozracivanje reaktorskih materijala po projektima VISA-2 i 3, 3. Osvajanje postupka izrade i ispitivanja kapsula i kenera VISA - I faza (I i II deo), II deo

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-02-15

    Experiments concerned with Projects VISA-2 and 3 demand construction of hermetization test-tubes, irradiation capsules, experimental devices and reactor channels as well as welding of fuel element claddings. For this purpose special materials as stainless steels, aluminium alloys, pure aluminium, magnesium, zirconium were chosen. these materials demand special procedure for welding. This report includes design and construction data with drawings of the special device for semiautomated circular welding.

  15. Operation and maintenance of the RB reactor, Annual report for 1977; Pogon i odrzavanje reaktora RB, Izvestaj o radu u 1977. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1977-07-01

    The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators.

  16. A model for reliability avaliation of the electrical supply source of the 1A3 and 1A4 control rods assemblies of Angra I reactor

    International Nuclear Information System (INIS)

    Yang, T.

    1978-01-01

    The reliability of the electrical power supply to the 4.16KV buses for the safety system operation of a nuclear power plant was studied. Particularly, Angra Unit I system was focused. Initially, reliability of each electrical supply source was estimated. Using a probabilistic approach based on the Markov processes, the system reliability was evaluated in terms of frequency and duration of loss of power supply and of the system failure probability evolution when one or more sources remained unavailable. Based on these results, certain reactor operating rules were proposed concerning later shutdown of the plant without compromising the nuclear reactor safety. A sensitivity analysis was also performed to show the different reliability parameter influences on final results. This analysis showed that the diesel system performs an important role in the power supply for a nuclear power plant [pt

  17. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA - Deo 1 - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1982. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M; Miokovic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1982-12-15

    outdated spare parts. Project concerning renewal of the reactor equipment was initiated during the past year according to the contract with the Soviet Atomenergoexport and IAEA which has planned to spend 1 000 000 of rubles for this project. [Serbo-Croat] Probni rad reaktora zapocet septembra 1981. godine na snazi od 2 MW sa 80% obogacenim gorivom nastavljen je u celoj 1982. prema prethodno napravljenom planu. Pocetno jezgro formirano ja sa 44 gorivna kanala sa po deset gorivnih elemenata. Prva polovina godine iskoriscena je za neophodna merenja i ispitivanja radnih parametara rektora i funkcionisanje sistema i opreme u radnim uslovima. U drugoj polovini godine zapocet je program probnog rada na visim snagama. Utvrdjeno je da ugradjeni visak reaktivnosti i kapacitet kontrolnih sipki zadovoljava sigurnosne kriterijume MAAE, ugradjeni visak reaktivnosti moze da omoguci rad na snazi od 4,7 MW u 4 mesecne kampanje sa po 15-20 dana rada, postoje povoljni uslovi za hladjenje jezgra pri pocetnoj konfiguraciji. Izmeren je efekat pocetnog zatrovanja na reaktivnost i raspodelu snage, izmerena je pocetna prostorna raspodela neutronskog fluksa koja iznosi 3,9 10{sup 13} cm{sup -2} s{sup -1} pri znazi od 2 MW. Odredjena je promena kalibracionog koeficijenta u sistemu za automatsko odrzavanje snage. Svi rezultati ukazuju da ce pri nominalnoj snazi od 4,7 Mw biti zadovoljeni svi kriterijum sigurnosti i postovana ogranicenja u odnosu na koriscenje goriva. Po dobijanju dozvole za rad na punoj snazi morace da se izvrsi dopunski probni rad na snagama od 3, 4, i 4,7 MW. Prelaz od pocetne konfiguracije sa 44 gorivna kanala u jezgru vrsice se postupno da bi se dostigla ravnotezna konfiguracija sa 72 gorivna kanala sa po 10 elemenata. Reaktor nije radio u septembu mesecu zbog radova na zameni dela cevovoda koji povezuje pumpnu stanicu na Dunavu sa horizontalnim taloznikom. Kontrola i odrzavanje opreme izvrsavani su redovno u granicama raspolozivosti rezerbinh delova. Teskocu pricinjava

  18. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Peixoto, Orpet J.M. [Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials, Avenida Rio Branco, 123/Grupo 515- Centro, CEP: 20040-005, Rio de Janeiro (Brazil); Diaz, Gustavo [National Regulatory Authority - Argentina, Av. Del Libertador 8250, (1429) Buenos Aires (Argentina)

    2015-07-01

    At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the

  19. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  20. Application of fuzzy logic in nuclear reactor control Part I: An assessment of state-of-the-art

    International Nuclear Information System (INIS)

    Herger, A.S.; Jamshidl, M.; Alang-Rashid, N.K.

    1995-01-01

    This article discusses the application of fuzzy logic to nuclear reactor control. The method has been suggested by many investigators in many control applications. Reviews of the application of fuzzy logic in process control are given by Tong and Sugeno. Because fuzzy logic control (FLC) provides a pathway for transforming human abstractions into the numerical domain, it has the potential to assist nuclear reactor operators in the control room. With this transformation, linguistically expressed control principles can be coded into the fuzzy controller rule base. Having acquired the skill of the operators, the FLC can assist an operator in controlling the complex system. The thrust of FLC is to derive a conceptual model of the control operation, without expressing the process as mathematical equations, to assist the human operator in interpreting incoming plant variables and arriving at a proper control action. To introduce the concept of FLC in nuclear reactor operation, an overview of the mythology and a review of its application in both nuclear and nonnuclear control application domains are presented along with subsequent discussion of fuzzy logic controllers, their structures, and their method of information processing. The article concludes with the application of a tunable FLC to a typical reactor control problem

  1. Job/task analysis for I ampersand C [Instrumentation and Controls] instrument technicians at the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Duke, L.L.

    1989-09-01

    To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs

  2. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

    2012-01-15

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  3. The training of the staff for work with radioactive materials and work on nuclear reactor in the Institute; Obuka kadrova za rukovanje radioizotopima i pogon nuklearnih reaktora u Institutu 'Boris Kidric' - Vinca

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Mladjenovic, O; Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    A short informational review of the activities in the 'Boris Kidric' Institute on the training courses for the use of radioactive materials and for operating nuclear reactors including power reactors. The survey of the courses is given in the enclosures. (author) Kratak informativni pregled delatnosti u IBK na kursevima za obuku kadrova u rukovanju readioaktivnim materijalima i pogonu nuklearnih reaktora, ukljucujuci reaktore snage. pregled kurseva i materijala za njih dati su u prilozima. (author)

  4. Radiological impact on the workers, members of the public, and environment from the partial decommissioning of Pakistan Research Reactor-I and its associated radioactive residues.

    Science.gov (United States)

    Ali, A; Orfi, S D; Manzur, H; Aslam, M

    2001-05-01

    The Pakistan Research Reactor-I (PARR-I) is a swimming pool type research reactor originally designed and built for a thermal power of 5 MW using High Enriched Uranium (HEU) fuel. In 1990-1991 the reactor was redesigned, partially decommissioned and recommissioned to operate with Low Enriched Uranium (LEU) fuel at a thermal power of 10 MW. An essential requirement, construction and commissioning of a wet spent fuel storage bay and fabrication of an irradiated fuel transfer cask were completed before actual dismantling of the reactor core. During the partial decommissioning operations, radioactive waste generated included 600 m3 low-level liquid radioactive waste and 14 m3 of solid radioactive waste with an average specific activity of 4.52 Bq ml(-1) and 2.22 kBq g(-1), respectively. External radiation doses of the workers were determined using TLD (NG 6,7) and direct reading dosimeters. The maximum individual external radiation dose received by any worker during this practice was 5 mSv, which was 25% of the annual dose limit of 20 mSv. Detection and measurement of internal contamination was carried out using bioassay techniques. During the whole operation, not a single case of internal contamination was detected. The ambient radiation levels around waste seepage pits are periodically monitored using TLD (G-2 cards) and G. M. radiation survey meters. Underground migration of radioactivity is checked by analyzing seepage water samples taken from boreholes that have been dug at different locations in the vicinity of the radioactive residues. The monitoring around disposal sites containing radioactive residues has been continued during the last 9 y and will be continued in the future. So far, no rise in the environmental gamma radiation dose level and migration of underground radionuclides has been found in the vicinity of these disposal sites. Working personal during the decommissioning of PARR-I have been found to be radiologically safe. Adherence to the ALARA

  5. RA reactor operation and maintenance in 1990 with comparative evaluation from 1986-1990, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1990. godini, uz uporedni pregled za period 1986-1990. godina

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Vasovic, B; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1990-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The new emergency cooling system and the reconstruction of the existing ventilation system were finished in 1989, the conditions for further reactor operation were fulfilled. In the meantime new licensing regulations adopted in 1988 were not demanding the mentioned conditions for reactors operated at power less than 10MW, RA reactor power being 6.5 MW. But the reactor could not be restarted due to planned renewal of the reactor instrumentation. It is planned to exchange the complete instrumentation by the end of 1991. Training program for the staff operating and maintaining the reactor components was prepared in 1985. Reconstruction, modification and construction of components demanded new documentation needed for further safe reactor operation. New version of RA reactor safety report was finished in 1986 according to the recommendations of IAEA and licensing regulations of Yugoslavia. In 1989, new documents were written covering regulations and instructions for reactor operation. The new reactor experimental loop was designed in 1986, and constructed and tested in 1990. All the reactor components were maintained by specific reactor services. Financing of the reactor remains a permanent problem. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. godine Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Radovi na izgradnji sistema za udesno hladjenje i rekonstrukciji postojeceg sistema specijalne ventilacije zavrseni su 1989. godine. Uslovi za nastavak rada reaktora

  6. Partial pressure (or fugacity) of carbon dioxide, dissolved inorganic carbon, pH, temperature, salinity and other variables collected from discrete sample and profile observations using CTD, bottle and other instruments from the EDWIN LINK in the North Atlantic Ocean and South Atlantic Ocean from 1996-04-15 to 1996-05-16 (NODC Accession 0113539)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NODC Accession 0113539 includes chemical, discrete sample, physical and profile data collected from EDWIN LINK in the North Atlantic Ocean and South Atlantic Ocean...

  7. Thermal power calibration of the TRIGA Mark I IPR-R1 reactor during the upgrading tests to 250 kW

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Maretti, Fausto Junior; Rezende, Hugo Cesar

    2002-01-01

    This paper presents the results and the methodology used to calibrate the thermal power of the TRIGA MARK I IPR-R1 Reactor in CDTN, Belo Horizonte, Brazil. This calibration was realized during the operation tests carried out to allow the reactor power upgrade from the current 100 kW to 250 kW. The methodology consisted in the measurement of the inlet and outlet temperature and the water flow in the primary cooling loop. The thermal balance together with the thermal losses gave the thermal power. There were made three sequences of tests. The first rising of the thermal power was made with the usual configuration of the core (59 fuel elements). After the changing of the ion chambers position and the control rod and the increase of the number of fuels (63 fuel elements), a new evaluation of the thermal power was accomplished, having been obtained a thermal power of 234 kW, for an indication of 250 kW in the lineal channel. After the return of the core to the initial configuration (59 fuel elements), it took place a new test, getting back the reactor to the power level of 100 kW. (author)

  8. The role of the IPR-R1 TRIGA Mark I research reactor in nuclear education and training in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Andrea V.; Mesquita, Amir Z.; Maretti Junior, Fausto; Souza, Rose Mary G.P.; Dalle, Hugo M.; Paiano, Silvestre, E-mail: avf@cdtn.br, E-mail: amir@cdtn.br, E-mail: fmj@cdtn.br, E-mail: souzarm@cdtn.br, E-mail: dallehm@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The revival of the Brazilian nuclear program has anticipated a large demand for training in nuclear technology. The Nuclear Technology Development Center (CDTN), a research institute of the Brazilian Nuclear Energy Commission (CNEN), offers the Operator Training Course on Research Reactors (CTORP). This course has existed since 1974 and about 258 workers were certificated by CTORP. This article describes the activities of CTORP and presents a proposal for its activities expansion in order to provide the current demand in the nuclear technology. Experimental research projects programs would be created in the postgraduate course at CDTN. In addition to the normal reactor physics topics addressed by CTORP, new subjects such as thermal hydraulic and instrumentation should be added and discussed too. (author)

  9. Numerical nodal simulation of the axial power distribution within nuclear reactors using a kinetics diffusion model. I

    International Nuclear Information System (INIS)

    Barros, R.C. de.

    1992-05-01

    Presented here is a new numerical nodal method for the simulation of the axial power distribution within nuclear reactors using the one-dimensional one speed kinetics diffusion model with one group of delayed neutron precursors. Our method is based on a spectral analysis of the nodal kinetics equations. These equations are obtained by integrating the original kinetics equations separately over a time step and over a spatial node, and then considering flat approximations for the forward difference terms. These flat approximations are the only approximations that are considered in the method. As a result, the spectral nodal method for space - time reactor kinetics generates numerical solutions for space independent problems or for time independent problems that are completely free from truncation errors. We show numerical results to illustrate the method's accuracy for coarse mesh calculations. (author)

  10. The role of the IPR-R1 TRIGA Mark I research reactor in nuclear education and training in Brazil

    International Nuclear Information System (INIS)

    Ferreira, Andrea V.; Mesquita, Amir Z.; Maretti Junior, Fausto; Souza, Rose Mary G.P.; Dalle, Hugo M.; Paiano, Silvestre

    2011-01-01

    The revival of the Brazilian nuclear program has anticipated a large demand for training in nuclear technology. The Nuclear Technology Development Center (CDTN), a research institute of the Brazilian Nuclear Energy Commission (CNEN), offers the Operator Training Course on Research Reactors (CTORP). This course has existed since 1974 and about 258 workers were certificated by CTORP. This article describes the activities of CTORP and presents a proposal for its activities expansion in order to provide the current demand in the nuclear technology. Experimental research projects programs would be created in the postgraduate course at CDTN. In addition to the normal reactor physics topics addressed by CTORP, new subjects such as thermal hydraulic and instrumentation should be added and discussed too. (author)

  11. Fission track dating method: I. Study of neutron flux uniformity in some irradiation positions of IEA-R1 reactor

    International Nuclear Information System (INIS)

    Osorio, A.M.; Hadler, J.C.; Iunes, P.J.; Paulo, S.R. de

    1993-06-01

    In order to use the fission track dating method the flux gradient was verified within the sample holder, in some irradiation positions of the IEA-R1 reactor at IPEN/CNEN, Sao Paulo. The fission track dating method considers only the thermal neutron fission tracks, to subtract the other contributions sample irradiations with a cadmium cover was performed. The neutron flux cadmium influence was studied. (author)

  12. Proceeding on the scientific meeting and presentation on basic research of nuclear science and technology (book I): physics, reactors

    International Nuclear Information System (INIS)

    Syarip; Prayitno; Samin; Agus Taftazani; Sudjatmoko; Pramudita Anggraita; Gede Sutresna W; Tjipto Sujitno; Slamet Santosa; Herry Poernomo; R Sukarsono; Prajitno

    2014-06-01

    Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is an annual activity held by Centre for Accelerator Science and Technology, National Nuclear Energy Agency, in Yogyakarta, for monitoring research activities achieved by the Agency. The papers presented in the meeting were collected into proceedings which were divided into two groups that are physics and nuclear reactors. The proceedings consists of three articles from keynote speakers and 25 articles from BATAN and others participants.(PPIKSN)

  13. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mattie, Patrick D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this study was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.

  14. Overview of the US stellarator reactor study

    International Nuclear Information System (INIS)

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1993-01-01

    This study, which uses a cost-minimization code that incorporates the ARIES costing and reactor component models with a I-D energy transport calculation, shows that a torsatron reactor could be competitive with a tokamak reactor

  15. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-02-15

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation. [Serbo-Croat] Pored osnovnih karakeristika reaktora RA, organizacije rada i finansijskih pokazatelja, razmatra se stanje opreme reaktora nakon 18 godina rada, pitanja dozvole za rad sa 80% obogacenim gorivom, problem skladistenja isluzenog goriva u bazenu zgrade reaktora i potreba za obnavljanjem komponenti opreme, pre svega elektronske.

  16. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  17. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    International Nuclear Information System (INIS)

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant

  18. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

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

  19. Reactor facility

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  20. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

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

  1. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

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

  2. Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

  3. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  4. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  5. Determination of average molecular weights on organic reactor coolants. I.- Freezing-point depression method for benzene solutions

    International Nuclear Information System (INIS)

    Carreira, M.

    1965-01-01

    As a working method for determination of changes in molecular mass that may occur by irradiation (pyrolytic-radiolytic decomposition) of polyphenyl reactor coolants, a cryoscopic technique has been developed which associated the basic simplicity of Beckman's method with some experimental refinements taken out of the equilibrium methods. A total of 18 runs were made on samples of napthalene, biphenyl, and the commercial mixtures OM-2 (Progil) and Santowax-R (Monsanto), with an average deviation from the theoretical molecular mass of 0.6%. (Author) 7 refs

  6. Nuclear power plants with reactors WWER-1000 type: today and tomorrow; AEhS s WWER-1000: nastoyashchee i budushchee

    Energy Technology Data Exchange (ETDEWEB)

    Molchanov, V; Biryukov, G; Novak, K [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    There are currently 19 NPP units based on WWER-1000 reactors working in Russia, Ukraine and Bulgaria. They are of four types: V-187, V-302, V-338, V-320. The design principles of these reactors comply with regulations of the eighties, and it is necessary to introduce improvements according to the new regulations and to the operation experience gained. Two approaches for safety and efficiency enhancement are described: AS-91 and AS-92. AS-91 implies gradual improvement of the base WWER-1000/V-320 design by incorporation of new design solutions avoiding the need of building large scale models. AS-92 refers to entirely new design which require experimental research by building a full scale models or by using natural stands. The latter approach will be used for NPP projects to be built after year 2000. The main new feature of AS-92 is the addition of passive safety systems to the active ones in order to protect the fuel from damage.

  7. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  8. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report; Naucnoistrazivacki projekt 'Istrazivacki nuclearni reaktor RB, (pogon i odrzavanje), Zavrsni elaborat projekta

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training.

  9. Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning; Prelaz toplote i pad pritiska reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem

    Energy Technology Data Exchange (ETDEWEB)

    Oka, S; Becirspahic, S [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1964-06-15

    Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning were investigated. Longitudinal and circumferential distributions of Sr-number of finnings in the fuel element are given. Dependences of St{sub kmin} and St{sub ksr} on the Re number are derived. The influence of gap between two fuel elements on the heat transfer, pressure drop is presented dependent on the Re number. The influence of mutual position of flow separators of two neighbouring fuel elements on the pressure drop and heat transfer is shown as well. Investigations were performed in the range of Re numbers from 15000 to 100000. Ispitivan je prelaz toplote i pad pritiska modela reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem. Dat je uzduznu i obimni raspored Sr-broja na orebrenju gorivnog elementa. Izvedene su zavisnosti St{sub kmin} i St{sub ksr} u funkciji od Re-broja. Pokazan je uticaj prekida izmedju dva gorivna elementa na prelaz toplote i pad pritiska u zavisnosti od Re-broja. Pokazan je uticaj medjusobnog polozaja razdeljivaca struje dva susedna gorivna elementa na pad pritiska i prelaz toplote. Ispitivanja su vrsena u oblasti Re-brojeva od 15000 do 100000 (author)

  10. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  11. Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-15

    This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

  12. Nokia pole ainus langev täht. RIM on hädas / Fredy-Edwin Esse

    Index Scriptorium Estoniae

    Esse, Fredy-Edwin

    2011-01-01

    Kanada tehnoloogiafirma Research In Motion kaotab ühe enam Põhja-Ameerikas turgu konkurentidele nagu Apple ja teised firmad, kes kasutavad Google'i Android operatsioonisüsteemi. Ka BlackBerry Playbook ei ole suutnud iPadiga konkureerida. Graafik

  13. The birth of plastic surgery: the story of nasal reconstruction from the Edwin Smith Papyrus to the twenty-first century.

    Science.gov (United States)

    Whitaker, Iain S; Karoo, Richard O; Spyrou, George; Fenton, Oliver M

    2007-07-01

    The nose is the central and most prominent feature on the human face; and on its shape, size, and appearance depends the relative facial beauty of the person. The objective of this article was to give a succinct and interesting account of the development of nasal reconstruction from antiquity to the present day. The authors present the story of nasal reconstruction, including those contributions not often cited in the English literature using articles sourced from MEDLINE, ancient manuscripts, original quotes, techniques, and illustrations. The story of rhinoplasty is one of peaks of achievement by individuals such as Sushruta, Branca, Tagliocozzi, Roe, and Joseph. Since Roe introduced the concept of cosmetic rhinoplasty, the evolution of nasal reconstructive techniques has reached such a level that the expectation is not only to restore form and function, but also to achieve excellent cosmetic appearance. Although repair of nasal injuries is the oldest form of reconstructive surgery, being cited in Egyptian papyrus inscriptions such as the Edwin Smith Papyrus dating back to 2500 to 3000 BC, its complexity continues to challenge surgeons today. This article is dedicated to those individuals who have devoted their lives and work to the advancement of the field of plastic surgery for the benefit of mankind.

  14. Software development of the mechanical vibration monitoring system of the CNA I reactor internals by neutron noise technique

    International Nuclear Information System (INIS)

    Wentzeis, Luis M.; Calvo, Maria D.

    2009-01-01

    The neutron noise analysis technique is an important predictive maintenance tool for early detection of failures such as sensor malfunctions and incipient mechanical problems located in the reactor internals. This technique was applied successfully in Argentina since 1987. The FER-GAEN group dependent of the CNEA developed the measuring system to detect anomalies as early as possible. The magnitude of interest in this analysis is the fluctuating component of the neutron flux known as 'neutron noise'. In order to improve and facilitate the analysis, a new software code was developed for the data acquisition of the neutron noise signals and neutron spectra estimation in the frequency domain. The RMS values related with the internals vibrations are calculated from these spectra and are chronologically displayed, in order to detect any anomalous vibration or incipient detector malfunction as early as possible. (author)

  15. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

    ... of Operation Under Severe Accident Conditions (Effective Immediately) I. The Licensees identified in... radioactive materials if the venting systems were used during severe accident conditions. The NRC staff..., 2012). Option 2 in SECY-12-0157 was to modify EA-12-050 to require severe accident capable vents (i.e...

  16. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  17. Radiation protection at the RA Reactor in 1989, Part -2, Decontamination, collection of treatment of fluid and solid radioactive waste, Annex 3; Deo 2 - Zastita od zracenja kod reaktora RA u 1989. godini, Dekontaminacija i intervencija, sakupljanje i obrada tecnih i cvrstih radioaktivnih otpadnih materija za potrebe reaktora RA - Prilog 3

    Energy Technology Data Exchange (ETDEWEB)

    Mandic, M; Vukovic, Z; Plecas, I; Knezevic, Lj; Lazic, S; Bacic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [Serbo-Croat] Tokom rada reaktora RA dolazi do stvaranja odredjenih cvrstih otpadnih materijala cija prosecna kolicina zavisi od vremena rada reaktora i aktivnosti koje se tamo obavljaju. Tokom remonta, kada reaktor ne radi kao i pri akcidentalnim situacijama nastaju vece kolicine otpadnih materijala koje zavise od obima i vrste remontnih operacija i obima dekontaminacije kontaminirane radne povrsine i kontaminiranog alata, predmeta, opreme, itd. Nastali otpadni materijali se razvrstavaju i pakuju na mestu nastanka prema odgovarajucim propisima u skladu sa principima zastite od zracenja i aspekta bezbednosti u cilju minimiziranja nepotrebnog ozracivanja ljudstva za preuzimanje, kontrolu, transport, naknadnu obradu RAO i dekontaminaciju. Pri nerutinskim operacijama (dekontaminacija, remont, kontaminiarni otpadni materijal velike zapremine i sl.), strucna sluzba Institita ZASTITA pruza strucne konsultacije i pomaze pri planiranju

  18. Study On Analytical Methods Of Tellurium Content In Natriiodide (Na131I) Radiopharmaceutical Solution Produced In The Dalat Nuclear Reactor

    International Nuclear Information System (INIS)

    Vo Thi Cam Hoa; Duong Van Dong; Nguyen Thi Thu; Chu Van Khoa

    2007-01-01

    This report describes the practical methods for analyzing of Tellurium content in Na 131 I solution produced at the Dalat Nuclear Research Institute. We studied analytical methods to control Tellurium content in final Na 131 I solution product used in medical purposes by three methods such as: spot test, gamma spectrometric and spectrophotometric methods. These investigation results are shown that the spot test method is suitable for controlling Tellurium trace in the final product. This spot test can be determinate Tellurium trace less than 10 ppm and are used to quality control of Na 131 I solution using in medical application. (author)

  19. Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation); Rezultati merenja zracenja u okolini i rezultati meteoroloskih merenja (materijal pripremljen radi dobijanja dozvole za pustanje Reaktora RA u probni rad)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-10-15

    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. [Serbo-Croat] Saglasno zahtevu za dobijanje dozvole za ponovno pustanje u probni rad reaktora RA, ovaj dokument sadrzi rezultate merenja zracenja u okolini (radnoj i zivotnoj) reaktora RA odnosno instituta 'Boris Kidric' kao i podatke o meteoroloskim merenjima. Sve merenja rade se prema programu mera zastite od zracenja koje se sprovode prakticno od prvog pustanja reaktora u rad krajem 1959. godine.

  20. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

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

  1. MLR reactor

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  2. H Reactor

    Data.gov (United States)

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

  3. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels (I-NERI Annual Report)

    International Nuclear Information System (INIS)

    Petti, David Andrew; Maki, John Thomas; Languille, Alain; Martin, Philippe; Ballinger, Ronald

    2002-01-01

    The objective of this INERI project is to develop improved fuel behavior models for gas reactor coated particle fuels and to develop improved coated-particle fuel designs that can be used reliably at very high burnups and potentially in fast gas-cooled reactors. Thermomechanical, thermophysical, and physiochemical material properties data were compiled by both the US and the French and preliminary assessments conducted. Comparison between U.S. and European data revealed many similarities and a few important differences. In all cases, the data needed for accurate fuel performance modeling of coated particle fuel at high burnup were lacking. The development of the INEEL fuel performance model, PARFUME, continued from earlier efforts. The statistical model being used to simulate the detailed finite element calculations is being upgraded and improved to allow for changes in fuel design attributes (e.g. thickness of layers, dimensions of kernel) as well as changes in important material properties to increase the flexibility of the code. In addition, modeling of other potentially important failure modes such as debonding and asphericity was started. A paper on the status of the model was presented at the HTR-2002 meeting in Petten, Netherlands in April 2002, and a paper on the statistical method was submitted to the Journal of Nuclear Material in September 2002. Benchmarking of the model against Japanese and an older DRAGON irradiation are planned. Preliminary calculations of the stresses in a coated particle have been calculated by the CEA using the ATLAS finite element model. This model and the material properties and constitutive relationships will be incorporated into a more general software platform termed Pleiades. Pleiades will be able to analyze different fuel forms at different scales (from particle to fuel body) and also handle the statistical variability in coated particle fuel. Diffusion couple experiments to study Ag and Pd transport through SiC were

  4. I-1 General; I-1 Opsti deo

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-07-01

    General part of the regulations and instructions for operation of the RA reactor includes fundamental data about the reactor properties, biological shields, purpose of reactor operation, organizational scheme, rights and responsibilities of the head of laboratory, leaders of the working teams and all the staff of the laboratory for RA reactor exploitation. Opsti deo knjige propisa i uputstva za rad reaktora sadrzi osnovne podatke o reaktoru, biloskoj zastiti, nameni reaktora, prava i duznosti nacelnika, sefova smena i osoblja laboratorije za eksploataciju reaktora RA.

  5. RA Reactor operation and maintenance (I-IX), Part II, Task 3.08/04; Pogon i odrzavanje reaktora RA (I-IX), II Deo, Zadatak 3.08/04

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

    Repair and maintenance work of the RA reactor components and equipment was longer than planned due to the following reasons: chemical decontamination of the reactor heavy water system indispensable for maintenance of the heavy water pumps; findings of contamination origin changed the maintenance schedule; some construction materials of the heavy water pumps which caused contamination had to be removed; a number of the planned operations had to be performed under increase level of gamma radiation which increased the time needed for repair. This report covers detailed description of all the maintenance and repair work done from 25 Jan to 22 Apr 1963.

  6. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I

    International Nuclear Information System (INIS)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes

  7. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

  8. Leis: aasta algus toob hinnatõusu. Balti turg kosub kriisist / Raivo Sormunen, Fredy-Edwin Esse ; kommenteerinud Kristo Oidermaa

    Index Scriptorium Estoniae

    Sormunen, Raivo, 1976-

    2011-01-01

    Toorainehindade tõusu tõttu vähenes IV kvartalis Premia Foodsi kasumlikkus ning kallinevad I kvartalis toodete hinnad. Premia Foodsi 2010. a. brutokasumist andis 54% jäätis, ülejäänu jagunes külmkauba- ja kalasegmendi vahel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-12-15

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

  10. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

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

  11. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  12. Study of the absorbers and gaps influence on the reactor reactivity, IZ-061-0071-1961; Ispitivanje uticaja apsorbera i supljina na reaktivnost reaktora, IZ-061-0071-1961

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    It has been foreseen by the contract to study theoretically and experimentally the influence of the absorbers and gaps on the reactivity of the reactor. Within theoretical study it was planned to develop a method and find the approximation methods for calculations of these effects. Experimental part include development of equipment and performing the experiment at the RB reactor. Since it has not been possible to perform the experiment due to lack of heavy water, only the theoretical part of the task was completed with additional theoretical study of the VISA-1 experimental loop. This report includes the following annexes: influence of absorbers and gaps on the reactivity of the reactor, and calculation of flux depression in the VISA-1 loop. [Serbo-Croat] Ugovor predvidja teorijske i eksperimentalne radove u vezi ispitivanja uticaja apsorbera i praznina na reaktivnost reaktora. Za teorijski deo predvidjena je razrada metode i nalazenje aproksimativnih metoda za proracun ovih efekata. Eksperimentalni deo predvidja pripremu uredjaj i izvodjenje eksperimenata na reaktoru RB. S obzirom na nedostatak teske vode i nemogucnost izvodjenja eksperimenata, zadatak je obavljan samo teorijski a dodata mu je i teorijska obrada petlje VISA-1. Zadatak sadrzi sledece priloge: uticaj apsorbera i supljina na reaktivnost reaktora i proracun depresije fluksa u petlji VISA-1.

  13. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex B - Report of the Technology Service; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1985. godini - Prilog B - Izvestaj o radu tehnoloske sluzbe

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Vukadin, Z; Stosic, O; Cupac, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1985-12-01

    Technology service is established according to the need of unified technology control of the reactor, based on the organizational structure developed in USA, a part of which was applied at the Krsko NPP. Tasks of this service are: regular control and recording of operating reactor parameters under steady state and accidental conditions; control of chemical state of the reactor core (control of Ph values and electro conductivity of the heavy water and technical water); dosimetry parameters control and recording; safety and optimisation analyses when planning reactor core modifications; (changing the type of fuel, design of new components); cooperation with the reactor users during isotope production, activation analyses, application of neutron beams; participating in research programs; planning refueling. The most important task performed by this service was conversion of the core from 2% to 80% enriched fuel elements. A methodology for optimizing the reactor application was developed. Operation of the heavy water purification system, heavy water parameters, and operation of the gas system were controlled. Radiation and contamination levels in the working environment were measured, individual and collective doses. [Serbo-Croat] Tehnoloska sluzba uspostavljena je u skladu sa ispoljenim potrebama jedinstvene tehnoloske kontrole reaktora, oslanjajajucui se na organizacionu semu koja je razvijena u SAD, a cija je jedna varijanta primenjena na NE Krsko. Zadaci sluzbe su obavljanje sledecih zadataka: redovna kontrola i evidencija radnih parametara reaktora u redovnim i akcidentalnim uslovima; kontrola hemijskog rezima u reaktoru (kontrola PH vrednosti i elektroprovodnosti teske i tehnicke vode); kontrola i evidencija tehnicke dozimetrije; sigurnosne i optimizacione analize prilikom planiranja izmena na reaktoru (promene vrste goriva, izgradnja novih elemenata opreme); saradnja sa korisnicima reaktora prilikom prozivodnje izotopa, aktivacione analize, koriscenja

  14. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  15. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  16. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  17. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-07-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels. [Serbo-Croat] Ovo (prvo) poglavlje sadrzi sledece: Opis reaktora RA; semu organizacije rada i rukovodjenja; prava i duznosti direktora i rukovodioca pogona reaktora, propise o rezimu rada i kretanja u zgradi reaktora, propise o izvodjenju eksperimenata, propise o unosenju uzoraka u eksperimentalne kanale reaktora.

  18. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  19. Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

    International Nuclear Information System (INIS)

    McCauley, E.W.; Pitts, J.H.

    1977-01-01

    The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

  20. Electromagnetic acoustic transducer (EMAT) defect characterization of nuclear reactor piping welds. Phase I. Final report, October 1985-March 1986

    International Nuclear Information System (INIS)

    Davis, T.J.; Thome, D.K.

    1986-05-01

    The Phase I workscope was successfully completed. This work was directed at determining the most promising methods for application of EMATs to stainless steel piping examination. It consisted of a literature review, evaluation of shear and longitudinal wave inspection modes, and evaluation of several signal processing techniques to enhance signal/noise ratios. The work involved both hardware and software development. A high degree of success was obtained during the course of the work, indicating that further exploitation of the technique is fully warranted. Defects as small as 0.1 cm deep could be detected in wrought stainless piping, and the ability to detect defects in thick centrifugally cast stainless samples was demonstrated. In addition, the techniques showed promise for sizing the flaws. These results were achieved through a combination of synthetic aperture processing, temporal averaging and low frequency illumination. Additional techniques were evaluated, including frequency analysis, angle beam scanning and multimode inspection, but were shown to be of limited benefit for the samples available in Phase I. However, these techniques may offer potential for discriminating between cracks and geometric reflectors. 56 refs., 21 figs

  1. Technical description of other types of reactors

    International Nuclear Information System (INIS)

    Vollmer, H.

    1977-01-01

    The paper reviews the development of reactor systems other than LWR, i. e. gas cooled reactors, heavy water reactors and fast breeders. The specific features of these reactors are discussed. Technical details on plant design of the various systems will be given as well as the present status-of-the-art. (orig.) [de

  2. The Unseen Déjà-Vu: From Erkki Huhtamo's Topoi to Ken Jacobs' Remakes: Commentary to Edwin Carels "Revisiting Tom Tom: Performative anamnesis and autonomous vision in Ken Jacobs' appropriations of Tom Tom the Piper's Son".

    Science.gov (United States)

    Strauven, Wanda

    2018-01-01

    This commentary on Edwin Carels' essay "Revisiting Tom Tom : Performative anamnesis and autonomous vision in Ken Jacobs' appropriations of Tom Tom the Piper's Son " broadens up the media-archaeological framework in which Carels places his text. Notions such as Huhtamo's topos and Zielinski's "deep time" are brought into the discussion in order to point out the difficulty to see what there is to see and to question the position of the viewer in front of experimental films like Tom Tom the Piper's Son and its remakes.

  3. (I) A Declarative Framework for ERP Systems(II) Reactors: A Data-Driven Programming Model for Distributed Applications

    DEFF Research Database (Denmark)

    Stefansen, Christian Oskar Erik

    This dissertation is a collection of six adapted research papers pertaining to two areas of research. (I) A Declarative Framework for ERP Systems: • POETS: Process-Oriented Event-driven Transaction Systems. The paper describes an ontological analysis of a small segment of the enterprise domain......, namely the general ledger and accounts receivable. The result is an event-based approach to designing ERP systems and an abstract-level sketch of the architecture. • Compositional Specification of Commercial Contracts. The paper describes the design, multiple semantics, and use of a domain....... • Using Soft Constraints to Guide Users in Flexible Business Process Management Systems. The paper shows how the inability of a process language to express soft constraints—constraints that can be violated occasionally, but are closely monitored—leads to a loss of intentional information in process...

  4. New Instruments and Principles for the Dimensional Measurement and Measurement of Spacing of Reactor Components; Nouveaux Instruments et Procedes de Mesure des Dimensions et de l'Espacement des Elements d'un Reacteur; Novye pribory i printsipy izmereniya razmerov i raspolozheniya komponentov reaktora; Nuevos Instrumentos y Principios para Medir las Dimensiones y la Separacion Entre Componentes de Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, P. [Institut Dr. Foerster, Reutlingen, Federal Republic of Germany (Germany)

    1965-09-15

    la practica. Se examina la medicion de recubrimientos no magneticos aplicados sobre materiales magneticos, se explican los principios de la medicion (metodos de campo magnetico con corriente continua y alterna) y se describen instrumentos para medir recubrimientos no magneticos de espesor comprendido entre 3 {mu}m y 20 mm. Se analiza el problema especial de medir depositos de estelita sobre paredes ferrnicas de recipientes de reactor. Se estudia la medicion de recubrimientos no conductores, aplicados sobre metales no ferrosos. Se explica el principio de medicion (corrientes de Foucault), se describe un instrumento destinado a este fin y se dan ejemplos tfpicos de. mediciones. El autor examina tambien la medicion de dimensiones ffsicas de los componentes metalicos de reactores, sin elementos en contacto con los mismos y describe diversos metodos aplicables a metales ferrosos y no ferrosos (metodo del campo magnetico con corriente continua y alterna, metodo de corrientes de Foucault), Se describen instrumentos y se dan ejemplos de mediciones a distancia del diametro, ovalidad, distorsion, etc., de componentes de reactor. Se exponen metodos para medir la separacion entre tales componentes, en la zona radiactiva del reactor. Se describe un instrumento para registrar perfiles de superficies y para indicacion directa de valores de aspereza ('Rauhtiefe', 'Glaettungstiefe', valor promedio y valor cuadratico medio). Se analizan ejemplos tfpicos del empleo de este instrumento para componentes de reactor. Se presta especial atencion a la posibilidad de usar un captador pequeno y de aplicaciones multiples mediante manipuladores en zonas y en materiales radiactivos. Se analiza el aumento de la aspereza superficial en funcion de la dosis de irradiacion. (author) [Russian] Full text: Rassmatrivaetsja izmerenie tolshhiny listov i tolshhiny stenok trub i kontejnerov iz austenitnyh i cvetnyh metallov. Obsuzhdajutsja dva metoda beskontaktnogo izmerenija tolshhiny stenok: a) metod

  5. Calorimeter measurements of absorbed doses at the heavy water enriched uranium reactor; Kalorimetrijska merenja apsorbovanih doza u reaktoru na tesku vodu i obogaceni uran

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Application of calorimetry measurements of absorbed doses was imposed by the need of good knowledge of the absorbed dose values in the reactor experimental channels. Other methods are considered less reliable. The work was done in two phases: calorimetry measurements at lower reactor power (13-80 kW) by isothermal calorimeter, and differential calorimeter constructions for measurements at higher power levels (up to 1 MW). This report includes the following four annexes, papers: Isothermal calorimeter for reactor radiation monitoring, to be published; Calorimeter dosimetry of reactor radiation, presented at the Symposium about nuclear fuel held in april 1961; Radiation dosimetry of the reactor RA at Vinca, published in the Bull. Inst. Nucl. Sci. 1961; Differential calorimeter for reactor radiation dosimetry.

  6. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Cupac, S.; Stanic, A.

    1990-01-01

    RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. 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 [sr

  7. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  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. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  10. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  11. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V.; Galvin, M.; Todreas, N.E.; Lombardi, C.V.; Maldari, F.; Ricotti, M.E.; Cinotti, L.

    2001-01-01

    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)

  12. Measurement of rod-drop time for control and safety rods at the RB reactor; Merenje vremena pada kontrolne sipke i sigurnosnih sipki reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Marinkovic, P; Stefanovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1990-12-15

    The objective of this task was to determine the thermal utilization factor f in a heterogeneous reactor cell. For this purpose reliable data about thermal neutron spatial distribution are needed. Spatial distribution of the thermal neutron flux in the reactor cell was measured by perturbation method which showed best results compared to other methods described in this paper experiments were done at the RB reactor.

  13. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I; Comparacion entre un modelo de diferencias finitas (PUMA) y uno de elementos finitos (DELFIN) para la simulacion del reactor de la CNA-I (central nuclear Atucha-I)

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C R [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs.

  14. Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

    Energy Technology Data Exchange (ETDEWEB)

    Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

    1962-04-15

    In search for semiconductors, which can be used in high-flux reactors in order to measure flux distributions, we irradiated SiC p-n junctions in the Belgium BR-1 reactor. Two types of SiC-diodes of different origin have been irradiated. These junctions are grown in the Lely-furnace. The change in forward and reverse characteristics have been measured during and after irradiation up to temperatures of 150{sup o}C, while measurements up to a temperature of 500{sup o}C are in progress. It has been found that one type resists BR-1 neutrons up to an integrated flux of 10{sup 15} n/cm{sup 2}, while the other resists irradiation up to a flux of 10{sup 17} n/cm{sup 2}. The changes in characteristics are given as well as the result of some annealing experiments. (author) [French] En recherchant des semi-conducteurs pouvant servir a mesurer les distributions de flux dans les reacteurs a haut flux de neutrons, les auteurs ont irradie des jonctions p-n au SiC dans le reacteur belge BR-1. Deux types de diodes a SiC d'origines differentes ont ete ainsi irradies. Les jonctions en question sont preparees par etirage dans le four Lely. Les auteurs ont mesure les modifications subies par les caracteristiques I-V apres et pendant l'irradiation a des temperatures allant jusqu'a 150{sup o}C; ils poursuivent leurs mesures dans la gamme des temperatures allant de 150{sup o}C a 500{sup o}C. Us ont constate que l'un des types de diode a SiC resiste aux neutrons du reacteur BR-1 jusqu'a 10{sup 15} n/cm{sup 2}, tandis que l'autre type resiste a l'irradiation jusqu'a 10{sup 17} n/cm{sup 2}. Les auteurs indiquent les modifications subies par les caracteristiques, ainsi que le resultat de certaines experiences de recuit. (author) [Spanish] Los autores estan tratando de encontrar semiconductores con los que sea posible medir distribuciones de flujo en reactores de flujo elevado, y con este fin irradiaron uniones p-n del SiC en el reactor BR-1 de Belgica. Irradiaron dos tipos de diodos de SiC de

  15. The experience of the fuel waste management of AM and BR-10 reactor facilities at SSC RF IPPE named after A.I. Leipunsky

    International Nuclear Information System (INIS)

    Kochetkov, L.A.; Mamaev, L.I.; Stuzhnev, Yu.A.

    1999-01-01

    8 research and experimental reactors have been created at the Institute industrial site. The majority of them have been or are being decommissioned now. During many decades the reactor of the first-in-the-world NPP -AM and the fast neutron reactor BR-5 (BR-10) are the main research reactor bases of the Institute. They have been in operation for 45 and 40 years respectively. At present the preparation work for their decommissioning is being carried out. One of the problems of that process is the fuel waste management which amount is about 13 tons. The possibility of its reprocessing is under consideration. (author)

  16. Reactor container

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  17. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

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

  18. 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

  19. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  20. Theoretical analysis of nuclear reactors (Phase II), I-V, Part V, Determining the fine neutron flux distribution by Pn method; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, V Deo, Odredjivanje fine raspodele fluksa Pn metodom

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-10-15

    Expression for spherical harmonic moments were applied. They were obtained by spherical harmonics expansion of monoenergetic transport equation. This report presents the procedure for calculating the neutron flux distribution in the nine-zone reactor cell of the RA reactor in Vinca. The procedure was modelled for digital computer ZUSE Z-23 by expansion of the diagram of the automated P{sub 3} code, which is adequate for P{sub n} code with minor changes. The needed subroutines were developed. The most important ones were those for modified first and second order Bessel functions of n-th order. Computer Z-23 was operating only 15 hours during three months, and thus only the subroutines for modified Bessel functions could be tested and the obtained results were excellent. For the mentioned reason the neutron flux distribution will be calculated in the forthcoming period. Koriscen je izraz za svernoharmonicne momente koji se dobija razvijanjem monoenergetske transportne jednacine u sverne harmonike. Dat je postupak za odredjivanje raspodele neutronskog fluksa u devet medijalnoj celiji reaktora RA u Vinci. Gornji postupak je logicki organizovan na digitalnoj masini ZUSE Z-23 razvijanjem tekuceg dijagrama automatskog P{sub 3} koda, koji sa malim izmenama odgovara Pn kodu. Razvijene su potrebne osnovne subrutine, medju kojima su najznacajnije one za modificirane Beselove funkcije prve i druge vrste n - tog reda. Masina ZUSE Z-23 bila je u radu svega 15 zasova u toku tri meseca, te sam stigao da testiram samo subrutine za modificirane Beselove funkcije koje su dale odlicne rezultate. Iz ovoga razloga nije se mogla dobiti trazena raspodela fluksa. To ce biti ucinjeno u narednom periodu posto cemo moci koristiti masinu Z-23 (author)

  1. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  2. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  3. Phase 1 study of metallic cask systems for spent fuel management from reactor to repository. Volume I. Phase 1 study summary

    International Nuclear Information System (INIS)

    1986-02-01

    It was proposed to perform a systems evaluation of metallic cask systems in order to define and examine the use of various metallic cask concepts or combination of concepts for the overall inventory management of spent fuel starting with its discharge from reactors to its emplacement in geologic repositories. This systems evaluation occurs in three phases. This three phase systems evaluation leads to a definition and recommendation of a sound and practical metallic cask system to accomplish efficient and effective management of spent fuel in the back end of the nuclear fuel cycle. Phase 1 Study objectives: establish system-wide functional criteria and assumptions; perform the systems engineering needed to define the metallic cask concepts and their feasibility; perform a screening evaluation of the technical and economic merits of the concepts; and recommend those to be included for a more detailed systems evaluation in Phase 2. Phase 2 Study objectives: refine the system-wide functional criteria and assumptions; perform the design engineering needed to enhance the validity and workability of those concepts recommended in Phase 1; and perform a more detailed systems evaluation. Phase 3 Study objectives: conclude the systems evaluation and develop an implementation plan. Volume I presents an overview of the detailed systems evaluation presented in Volume II

  4. Szilard-Chalmers effect in solid H I O{sub 4}. 2 H{sub 2} O by neutron irradiation (source-reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Takriti, S [radiochemistry laboratory, syrian atomic energy commission P.O. Box 6091 Damascus, (Syrian Arab Republic)

    1995-10-01

    The Szilard-Chalmers effect in solid periodic acid was investigated. In order to study the initial distribution of {sup 128I} o{sub 4} as a function of neutron flux, samples were irradiated utilizing both neutron source ({sup 241} Am-Be), the manual vertical irradiation channel and the thermal column of ET-R R-1 research reactor in Egypt. The initial retention reached a maximum of 40% after 120 minutes at 5.5 x 10 {sup 8} n s{sup -1} cm {sup -2}. The data was analysed using first order reaction. As a result, the activation Ko= 2.82 x 10 {sup 11} (S{sup -1}), respectively. Kinetics comparison of the dehydration and irradiation reactions for this solid showed disorder in the crystallographic form. Such disorder may be the result of dehydration or irradiation reactions, where the loss of water molecule will lead to formation of vacancies which, in turn, are responsible for the distribution process. 6 figs., 1 tab.

  5. Control of the working environment dosimetry and technical radiation protection at the RA reactor, Part I; Deo I: Kontrola radne sredine - poslovi dozimetrije i tehnicke zastite od zracenja kod reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, N; Bjelanovic, J; Minicic, Z; Komatina, R; Raicevic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted to 16.9 mSV during past 10 months. Individual exposures for 9/10 of the personnel were less than 1/10 of the permissible annual exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. The last part analyzes accidents occurred at the reactor during 1984. It was found that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel.

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  7. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

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

  8. Research nuclear reactors and their role in nuclear-power program; Istrazivacki nuklearni reaktori i njihova uloga u nuklearno-energetskom programu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1980-07-01

    This paper deals with the role of experimental and research reactors in the nuclear power program. In addition to the overall analysis it contains more detailed description of experimental possibilities and operation properties of reactors RA, RB in Vinca, Belgrade and TRIGA in Ljubljana.

  9. Calculation of physical and thermo hydro-dynamic parameters of a thermal research reactor; Prorachun fizichkih i toplotno hidro-dinamichkih parametara termichkog istrazhivachkog reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M; Spasojevic, D; Jovic, V; Marinkovic, N [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    The paper presents initial activities on creating a design concept of a new thermal research reactor, which should be built according to the research and development program in the field of nuclear fuel cycle technologies. For one possible type of such a reactor basic design parameters are specified and some preliminary results of nuclear, thermal and hydrodynamic design calculations are given. (author)

  10. RA Research nuclear reactor, Part 1, RA reactor operation and maintenance in 1993, with comparative review for the period 1991 - 1993, Annex 3; Projekat Istrazivacki nuklearni reaktor RA - 1 Deo Pogon i odrzavanje nuklearnog reaktora RA u 1993. godini, uz uporedni pregled za period 1991 - 1993. - prilog 3

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1993-12-15

    RA reactor was not operated during 1993 because of the complete instrumentation exchange. Although it has been planned to exchange the complete instrumentation until the end of 1993, and to start reactor operation in the first half of 1993 this was not fulfilled because the instrumentation was not delivered until the end of 1993. Main activities during past seven years were related to construction of the emergency cooling system; repair and reconstruction of the system for handling the spent fuel and improvement of spent fuel storage conditions; exchange of the aged instrumentation. Other reactor components and systems, reactor core, primary coolant loop and gas circulation system are in good condition concerning future start-up. [Serbo-Croat] U 1993. godini reaktor nije bio u pogonu zbog zamene njegove celokupne instrumentacije. Iako je bilo planirano da se celokupna instrumentacija zameni do kraja 1993. te da reaktor pocne sa radom u prvoj polovini 1993. Ovo nije ispunjeno jer celokupna oprema nije isporucena ni do kraja 1993. godine. Osnovni zahvati koji su u proteklih sedma godina izvrseni, odnosili su se na izgradnju sistema za udesno hladjenje, rekonstrukciju sistema za rukovanje ozracenim gorivom i poboljsanje uslova za stokiranje ovog goriva, zamenu instrumentacije. Ostali sistemi reaktora, reaktorsko jezgro, primarno kolo hladjenja i sistem za cirkulaciju gasa su u dobrom stanju i mogu se nesmetano koristiti u buducem radu.

  11. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

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

  12. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  13. Operational report, Advantages of gradual introducing of highly enriched fuel into the RA reactor core from economic aspect and users needs; Radni izvestaj, Prednost postupka parcijalnog uvodjenja visokoobogacenog goriva u reaktor RA sa aspekta ekonomicnosti i potreba korisnika

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-10-14

    The possibility of increasing the neutron flux in the RA reactor was considered for a number of years. The possibilities of reactor reconstruction are not realistic and they should be disregarded. The possibility that remains is to achieve higher neutron flux by improving the fueling scheme and above all by introducing highly enriched fuel into the reactor core. Decision to purchase highly enriched fuel was quicker due to the fact that the 2% enriched uranium fuel is not fabricated any more. There are two procedures for exchanging the fuel in the reactor core: a) removal of partially spent 2% enriched fuel and formation of the core with fresh highly enriched fuel; b) gradually introducing the new fuel into the existing RA reactor core according to a special transfer regime. This report includes some comparative analyses of these two procedures from both economic point of view and the needs of users, as well as some technical conditions. These results are in favour of gradual introducing of new fuel into the reactor core. relevant direct savings amount to 3 000 000 dinars. Some of the most important advantages cannot be estimated in this way. This report does not cover the safety analyses results which are presented in a series of other papers. [Serbo-Croat] Vec vise godina razmatra se mogucnost za povecanje neutronskog fluksa u reaktoru RA. Mogucnosti za rekonstrukciju reaktora RA u tom smislu su minimalne i realno ih treba odbaciti. Prema tome preostaje da se povecanje neutronskog fluksa postigne usavrsavanjem seme izmene goriva, a pre svega uvodjenjem goriva sa visokim stepenom obogacenja u reaktor RA. Donosenje odluke o nabavci visokoobogacenog goriva i njegovom uvodjenju u reaktor ubrzano je i cinjenicom da se staro 2% obogaceno uransko gorivo vise ne proizvodi. Postoje dva postupka za prevodjenje reaktora na ovo gorivo: a) Uklanjanjem poluistrosenog 2% obogacenog goriva iz reaktora i formiranjem jezgra iskljucivo od svezeg visokoobogacenog goriva, b

  14. Radiation protection at the RA reactor in 1987, Part I: Control of the working environment - dosimetry and radiation protection at the RA reactor, Annex 1; Prilog 1, Zastita od zracenja kod reaktora RA u 1987. godini - Deo 1: Kontrola radne sredine - poslovi dozimetrije i tehnicke zastite od zracenja kod reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M; Bjelanovic, J; Minincic, Z; Komatina, R; Raicevic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Laboratory for radiation and environmental proetecion, Beograd (Serbia and Montenegro)

    1987-12-15

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted was less than 6.0 mSv during past 10 months. Individual exposures for 9/10 of the personnel were less than 1/10 of the annual permissible exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. During 1987 there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel. [Serbo-Croat] U ovom izvestaju prikazani su analizirani reprezentativni rezultati sakupljeni u okviru kontrole radne sredine i tehnicke zastite od zracenja reaktora RA. U prvom delu izvestaja izlozeni su podaci o osnovnim vidovima izlaganja zracenju i statisticki pregled ukupnog broja radiacionih merenja. Dati su takodje rezultati merenja sadrzaja radioktivnih gasova i aerosola u vazduhu, kao i stepena kontaminacije povrsina, odece i otkrivenih delova tela osoblja. U drugom delu izvestaja izlozeni su rezultati analize ozracivanja radnog osoblja. Utvrdjeno je da je maksimalna individualna doza spoljasnjeg izlaganja u proteklih 10 meseci bila 6,0 mSv, a da su pojadinacna izlaganja vise od 9/10 radnog osoblja bila manja od 1/10 godisnje granicne vrednosti. Dati su takodje uporedni podaci o ozracivanju osoblja u prethodnoj, kao i u pet proteklih godina, iz kojih

  15. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  16. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  17. ONDERHOUD MET EDWIN ARRISON1

    African Journals Online (AJOL)

    So ons het daardie week gehad, en in daardie week het daar 'n hele klompie goed gebeur waarvan ek heeltemal onbewus was. Byvoorbeeld, ek het skielik daardie week besef iemand soos Jeremiah. Wright kom na die konferensie toe. Ek weet nie of jy vir Jeremiah Wright ken nie, maar hy was Barack Obama se pastoor ...

  18. Edwin M. McMillan

    Science.gov (United States)

    , Lynne Annette, David Michael, and Stephen Keith. Dr. Glenn Seaborg with Ion-Exchanger illusion column of (1947- 1951); William H. Nichols Medal (New York Section of the American Chemical Society, 1948); member Medical School; they have three children: Ann Bradford, David Mattison, and Stephen Walker. He attended

  19. Twenty years of chemistry associated with the needs and utilization of nuclear reactors at the 'Boris Kidric' Institute of nuclear sciences, Vinca, Yugoslavia; Dvadeset godina hemije vezane za potrebe i koriscenje nuklearnih reaktora u Institutu za nuklearne nauke 'Boris kidric' i Vinci

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-07-01

    This publication covers nine review papers on the following topics related to the needs and utilization of nuclear reactors in the Boris Kidric Institute of nuclear sciences during previous twenty years: radiochemistry, hot atom chemistry, isotope production, spent nuclear fuel reprocessing, chemistry of transuranium elements; liquid radioactive waste processing, purification of reactor coolant water by inorganic ion exchangers, research related to deuterium concentration processes, and chemical dosimetry at the RA reactor. [Serbo-Croat] Ova publikacija obuhvata devet radova, po sledecim naslovima, a odnose se na potrebe i uslove nuklearnih reaktora u Institutu za nuklearne nauke 'Boris Kidric' tokom prethodnih dvadeset godina: radiohemija, hemija vruceg atoma, proizvodnja radioaktivnih izotopa, prerada isluzenog nuklearnog goriva, hemija transuranskih elemenata, obrada radioaktivnih otpadnih voda, preciscavanje vode za hladjenje nuklearnih reaktora pomocu neorganskih jonoizmenjivaca, istrazivanje procesa za koncentrovanje deuterijuma i hemijska dozimetrija reaktora RA.

  20. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  1. A review of calculation methods for fast and intermediate reactors; Expose des methodes pour le calcul de reacteurs a neutrons rapides et intermediaires; Obzor metodov rascheta reaktorov na promezhutochnykh i bystrykh nejtronakh; Estudio panoramico de los metodos de calculo de los reactores rapidos e intermedios

    Energy Technology Data Exchange (ETDEWEB)

    Marchuk, G I [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

    1962-03-15

    de ecuaciones fundamentales y conjugadas de la teoria de los multigrupos. Expone luego diversas aplicaciones de la teoria de la perturbacion a los problemas del calculo fisico del reactor. Examina los metodos numericos de resolucion de las ecuaciones fundamentales y conjugadas que expresan el funcionamiento del reactor sobre la base del metodo de los armonicos esfericos. Explica asimismo como se utiliza el metodo de las caracteristicas en la solucion de problemas relativos a la masa critica del reactor. Describe los metodos de calculo de los reactores con moderadores que contienen hidrogeno y, por fin, expone las bases de un modelo efectivo fundado en la teoria de un solo grupo, aplicable al reactor. (author) [Russian] Obsuzhdaetsya razvitie metodov rascheta yadernykh reaktorov na promezhutochnykh i bystrykh nejtronakh. Rassmatrivayuts ya razlichnye postanovki zadach fizicheskogo rascheta. Obsuzhdaetsya uchet rezonansnykh ehffektov. Vvodyatsya v rassmotrenie mnogogruppovy e sistemy 'osnovnykh i sopryazhennykh uravnenij. Daetsya razlichnoe primenenie teorii vozmushchenij k zadacham fizicheskogo rascheta reaktora. Rassmatrivayuts ya chislennye metody resheniya osnovnykh i sopryazhennykh uravnenij reaktora v priblizhenii metoda sfericheskikh garmonik. Daetsya primenenie metoda kharakteristik k resheniyu zadach na kriticheskuyu massu reaktora. Izlagayutsya metody rascheta reaktorov s vodorodsoderzhashchim i zamedlitelyami . Izlagayutsya osnovy ehffektivnoj odnogruppovoj modeli reaktora. (author)

  2. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    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.)

  3. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  4. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  5. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-15

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

  7. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I

    International Nuclear Information System (INIS)

    Grant, C.R.

    1996-01-01

    The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs

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

    Data.gov (United States)

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

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

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

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

  10. New start-up channels and multichannel analyzer at the RB reactor; Novi start-up kanali i videkanalni analizator na reaktoru Rb

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Markovic, H; Vranic, S; Dimitrijevic, Z; Pesic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1978-01-15

    New start-up channels and a multichannel analyzer were purchased in 1977 for the RB reactor. Both start-up channels contain BF{sub 3} neutron detectors, preamplifier, amplifier, single-channel analyzer, scaler, ratemeter, control unit, recording instrument. This document contains detailed technical description of these devices as well as characteristics of the multichannel analyzer which is being tested and will be used for measuring irradiation in the vicinity of the reactor.

  11. Emergency planning of the city of Munich with reference to nuclear facilities, especially the nuclear power stations Isar I and II, resp. the reactor in Garching

    International Nuclear Information System (INIS)

    1990-01-01

    During the hearing of Munich's city council of 13.7.1990 thirteen experts were heard on the following subjects: Hazard potential of Isar reactors and FRM reactor and appropriate radioactive waste transports; responsibilities in emergency planning. Some of the experts cannot visualize a major accident and propose not to cater for it. Shelters and evacuation are not planned for Munich, both solutions not being realizable for all inhabitants. Nuclear phaseout is seen by some as a measure of prevention. (HSCH) [de

  12. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David [Idaho National Lab. (INL), Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab. (INEEL); Martin, Philippe [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Phelip, Mayeul [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Ballinger, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2004-12-01

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  13. Kinetic modeling of cellulosic biomass to ethanol via simultaneous saccharification and fermentation: Part I. Accommodation of intermittent feeding and analysis of staged reactors.

    Science.gov (United States)

    Shao, Xiongjun; Lynd, Lee; Wyman, Charles; Bakker, André

    2009-01-01

    The model of South et al. [South et al. (1995) Enzyme Microb Technol 17(9): 797-803] for simultaneous saccharification of fermentation of cellulosic biomass is extended and modified to accommodate intermittent feeding of substrate and enzyme, cascade reactor configurations, and to be more computationally efficient. A dynamic enzyme adsorption model is found to be much more computationally efficient than the equilibrium model used previously, thus increasing the feasibility of incorporating the kinetic model in a computational fluid dynamic framework in the future. For continuous or discretely fed reactors, it is necessary to use particle conversion in conversion-dependent hydrolysis rate laws rather than reactor conversion. Whereas reactor conversion decreases due to both reaction and exit of particles from the reactor, particle conversion decreases due to reaction only. Using the modified models, it is predicted that cellulose conversion increases with decreasing feeding frequency (feedings per residence time, f). A computationally efficient strategy for modeling cascade reactors involving a modified rate constant is shown to give equivalent results relative to an exhaustive approach considering the distribution of particles in each successive fermenter.

  14. Study of the boron homogenizing process employing an experimental low-pressure bench simulating the IRIS reactor pressurizer – Part I

    International Nuclear Information System (INIS)

    Bezerra, Jair de Lima; Lira, Carlos Alberto Brayner de Oliveira; Barroso, Antonio Carlos de Oliveira; Lima, Fernando Roberto de Andrade; Bezerra da Silva, Mário Augusto

    2013-01-01

    Highlights: ► Experimental bench with test section made of transparent acrylic, simulating the pressurizer reactor IRIS. ► Workbench used to study the process of homogenization of boron in the pressurizer IRIS nuclear reactor. ► Results were obtained through videos and digital photos of the test section. - Abstract: The reactivity control of a nuclear reactor to pressurized water is made by means of controlling bars or by boron dilution in the water from the coolant of a primary circuit. The control with boron dilution has great importance, despite inserting small variations in the reactivity in the reactor, as it does not significantly affect the distribution of the neutron flux. A simplified experimental bench with a test section manufactured in transparent acrylic, was built in reduced scale as to be used in a boron homogenizing process, simulating an IRIS reactor pressurizer (International Reactor Innovative and Secure). The bench was assembled in the Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE), an entity linked to the Comissão Nacional de Energia Nuclear (CNEN), Recife – PE

  15. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  17. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

    Directory of Open Access Journals (Sweden)

    Z. Gholamzadeh

    2015-12-01

    Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  20. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  1. Breazeale Reactor Modernization Program

    International Nuclear Information System (INIS)

    Davison, C. C.

    2003-01-01

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future

  2. Gasification in pulverized coal flames. Final report (Part I). Pulverized coal combustion and gasification in a cyclone reactor: experiment and model

    Energy Technology Data Exchange (ETDEWEB)

    Barnhart, J. S.; Laurendeau, N. M.

    1979-05-01

    A unified experimental and analytical study of pulverized coal combustion and low-BTU gasification in an atmospheric cyclone reactor was performed. Experimental results include several series of coal combustion tests and a coal gasification test carried out via fuel-rich combustion without steam addition. Reactor stability was excellent over a range of equivalence ratios from .67 to 2.4 and air flowrates from 60 to 220 lb/hr. Typical carbon efficiencies were 95% for air-rich and stoichiometric tests and 80% for gasification tests. The best gasification results were achieved at an equivalence ratio of 2.0, where the carbon, cold gas and hot gas efficiencies were 83, 45 and 75%, respectively. The corresponding product gas heating value was 70 BTU/scf. A macroscopic model of coal combustion in the cyclone has been developed. Fuel-rich gasification can also be modeled through a gas-phase equilibrium treatment. Fluid mechanics are modeled by a particle force balance and a series combination of a perfectly stirred reactor and a plug flow reactor. Kinetic treatments of coal pyrolysis, char oxidation and carbon monoxide oxidation are included. Gas composition and temperature are checked against equilibrium values. The model predicts carbon efficiency, gas composition and temperature and reactor heat loss; gasification parameters, such as cold and hot gas efficiency and make gas heating value, are calculated for fuel-rich conditions. Good agreement exists between experiment and theory for conditions of this investigation.

  3. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

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

  4. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-02-15

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

  6. The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

    Energy Technology Data Exchange (ETDEWEB)

    Heineman, R. E. [General Electric Company, Richland, WA (United States)

    1964-02-15

    introduccion de cambios en el manejo de los reactores ya existentes, como instrumentos de investigacion en la esfera de la fisica de los reactores y como medio de ensenanza. Compara tambien los capitales invertidos en esas instalaciones y los gastos de explotacion. Describe el perfeccionamiento de nuevas tecnicas experimentales que estas instalaciones permiten aplicar con miras a satisfacer la demanda de nuevos datos experimentales. Es menester tener presentes todos estos datos para poder predecir la evolucion de las necesidades y las tendencias futuras en el empleo de estas instalaciones para los estudios de los reactores de potencia. La memoria describe sucintamente el reactor para el estudio de constantes fisicas e indica la manera en que se piensa utilizarlo en el marco de esa evolucion. (author) [Russian] V Hjen- fordskih laboratorijah v techenie pochti 15 let provodjatsja jeksponencial'nye reaktornye iz- merenija na grafito-uranovyh reshetkah. Hotja rezul'taty jetih opytov ispol'zovalis' dlja opredelenija laplasianov predlagaemyh proizvodjashhih reaktorov, oni takzhe sodejstvovali razvitiju ponimanija fiziki reaktorov jetih sistem. Davno priznano, chto poleznost' kri- ticheskogo opyta ogranichena vvidu ego bol'shogo masshtaba i nedostatochnoj chuvstvitel'nosti v otnoshenii nebol'shih lokalizovannyh narushenij sistemy. Zatem mysl' byla napravlena na sozdanie cel'nogo opytnogo reaktora, v kotorom bylo by svedeno do minimuma kolichest- vo materialov, neobhodimyh dlja poluchenija nuzhnyh dannyh. Jeta popytka privela k postrojke usovershenstvovannoj kriticheskoj ustanovki s neskol'kimi zonami reaktora dlja izmerenija fizicheskih konstant PCTR. Ustanovka ispol'zuetsja dlja okazanija sodejstvija pri razrabot- ke proekta po fizike reaktorov dlja neskol'kih jenergeticheskih reaktorov. Krome togo,re- aktor RSTNjavljaetsja ustanovkoj obshhego naznachenija dlja provedenija izmerenij poperechnyh sechenij na reaktore i dlja opredelenija differencial'nyh i integral'nyh fizicheskih para

  7. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  8. Oklo natural reactor

    International Nuclear Information System (INIS)

    Fujii, Isao

    1985-01-01

    In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

  9. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

  10. Reactor safety

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

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

  13. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

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

  14. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

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

  15. The effective lifetime and temperature coefficient in a coupled fast-thermal reactor; Temps de vie effectif et coefficient de temperature dans un reacteur a couplage neutrons rapides-neutrons thermiques; Ehffektivnyj srok zhizni i temperaturnyj koehffitsient nejtronov v dvoyakom reaktore na bystrykh i teplovykh nejtronakh; Vida efectiva y coeficiente de temperatura en un reactor con acoplamiento rapido-termico

    Energy Technology Data Exchange (ETDEWEB)

    Haefele, W. [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    thermique. Le probleme est etudie avec un groupe de neutrons retardes (au sens habituel). Un formalisme exprime le temps de vie effectif et le coefficient de temperature aux differents stades de la saute de puissance. L'auteur indique les sautes de puissance pour differentes valeurs de {alpha}{sub 0} jusqu'a ce que soit atteinte la limite de la cinetique du reacteur a neutrons rapides. (author) [Spanish] La teoria de los sistemas acoplados fue ampliamente desarrollada por Avery y sus colaboradores en el Argonne National Laboratory. Una de las caracteristicas mas interesantes de los sistemas acoplados es la prolongacion de la vida efectiva de los neutrones. La componente termica actua como una especie de retardador neutronico. Como en la teoria de los neutrones retardados, el efecto retardador desaparece cuando la reactividad adquiere un valor suficientemente elevado para que la componente rapida alcance la criticidad independientemente . El autor examina un reactor con acoplamiento en el que la componente rapida sufre un salto instantaneo de reactividad {alpha}{sub 0}. La temperatura aumenta como consecuencia del incremento del nivel de potencia y comienzan a actuar dos coeficiente termicos: el que corresponde a la componente rapida y el coeficiente de temperatura de la componente termica. El problema se estudia en relacion con un grupo de neutrones retardados (en el sentido corriente del termino). El autor presenta una serie de formulas que expresan la vida efectiva de los neutrones y el coeficiente de temperatura en las diferentes etapas del salto de reactividad. El autor indica esos saltos para distintos valores de {alpha}{sub 0}, basta alcanzar el limite correspondiente a la cinetica de los reactores de neutrones rapidos. (author) [Russian] Teoriya dvoyakikh sistem byla podrobno razrabotana R. Ehjveri i sotrudnikami v Argonskoj natsional'noj laboratorii. Odnim iz osnovnykh interesnykh momentov v sparennoj sisteme yavlyaetsya bol'shij ehffektivnyj srok zhizni nejtronov

  16. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

  18. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

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

  19. A Case Study: Implementation of a Management System for the TRIGA Mark II Research Reactor at the Laboratory of Applied Nuclear Energy (LENA) of the University of Pavia, Italy. Annex I

    International Nuclear Information System (INIS)

    2013-01-01

    This annex provides an example for the implementation of a management system for operating organizations of research reactors, based on a case study in which the implementation of such a system has been completed. The case study relates the experience of the Applied Nuclear Energy Laboratory (hereafter referred to as LENA) of the University of Pavia, Italy. This example is used because of the recent completion of the implementation of an integrated management system, and also because of the specific characteristics of the organization (such as the limited number of staff, limited financial resources, etc.), which are often typical for organizations that operate smaller research reactors. Section I-1 gives a brief presentation of the organization, including the scope of work, the main activities performed, the organizational structure, the identification of interested parties and the applicable requirements and standards. Section I-2 describes the LENA Management System, the reasons for its implementation, the stages of its development and the processes involved. Some practical examples related to the development of the LENA Management System are discussed in Section I-3, indicating the choices made by the organization. In particular, Section I-3.12 shows the correlation between the LENA Management System processes and the processes considered in the main body of this publication.

  20. Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part I: Theoretical Considerations

    Directory of Open Access Journals (Sweden)

    Éric Dumont

    2017-03-01

    Full Text Available This study demonstrated that a laboratory-scale recirculation closed-loop reactor can be an efficient technique for the determination of the Clean Air Delivery Rate (CADR of PhotoCatalytic Oxidation (PCO air purification devices. The recirculation closed-loop reactor was modeled by associating equations related to two ideal reactors: one is a perfectly mixed reservoir and the other is a plug flow system corresponding to the PCO device itself. Based on the assumption that the ratio between the residence time in the PCO device and the residence time in the reservoir τP/τR tends to 0, the model highlights that a lab closed-loop reactor can be a suitable technique for the determination of the efficiency of PCO devices. Moreover, if the single-pass removal efficiency is lower than 5% of the treated flow rate, the decrease in the pollutant concentration over time can be characterized by a first-order decay model in which the time constant is proportional to the CADR. The limits of the model are examined and reported in terms of operating conditions (experiment duration, ratio of residence times, and flow rate ranges.