Technical evaluation report on the proposed design modifications and technical-specification changes on grid voltage degradation for the San Onofre Nuclear Genetating Station, Unit 1

International Nuclear Information System (INIS)

Selan, J.C.

1982-01-01

This report documents the technical evaluation of the proposed design modifications and Technical Specification changes for protection of Class 1E equipment from grid voltage degradation for the San Onofre Nuclear Generating Station, Unit 1. The review criteria are based on several IEEE standards and the Code of Federal Regulations. The evaluation finds that the proposed design modifications and Technical Specification changes will ensure that the Class 1E equipment will be protected from sustained voltage degradation

Continental Shelf Morphology and Stratigraphy Offshore San Onofre, CA: The Interplay Between Rates of Eustatic Change and Sediment Supply

Science.gov (United States)

Klotsko, Shannon; Driscoll, Neal W.; Kent, Graham; Brothers, Daniel

2016-01-01

New high-resolution CHIRP seismic data acquired offshore San Onofre, southern California reveal that shelf sediment distribution and thickness are primarily controlled by eustatic sea level rise and sediment supply. Throughout the majority of the study region, a prominent abrasion platform and associated shoreline cutoff are observed in the subsurface from ~ 72 to 53 m below present sea level. These erosional features appear to have formed between Melt Water Pulse 1A and Melt Water Pulse 1B, when the rate of sea-level rise was lower. There are three distinct sedimentary units mapped above a regional angular unconformity interpreted to be the Holocene transgressive surface in the seismic data. Unit I, the deepest unit, is interpreted as a lag deposit that infills a topographic low associated with an abrasion platform. Unit I thins seaward by downlap and pinches out landward against the shoreline cutoff. Unit II is a mid-shelf lag deposit formed from shallower eroded material and thins seaward by downlap and landward by onlap. The youngest, Unit III, is interpreted to represent modern sediment deposition. Faults in the study area do not appear to offset the transgressive surface. The Newport Inglewood/Rose Canyon fault system is active in other regions to the south (e.g., La Jolla) where it offsets the transgressive surface and creates seafloor relief. Several shoals observed along the transgressive surface could record minor deformation due to fault activity in the study area. Nevertheless, our preferred interpretation is that the shoals are regions more resistant to erosion during marine transgression. The Cristianitos fault zone also causes a shoaling of the transgressive surface. This may be from resistant antecedent topography due to an early phase of compression on the fault. The Cristianitos fault zone was previously defined as a down-to-the-north normal fault, but the folding and faulting architecture imaged in the CHIRP data are more consistent with a

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

Variabilidad fisicoquímica del agua en la ciénaga El Eneal, reserva natural Sanguaré municipio de San Onofre Sucre, Colombia

Directory of Open Access Journals (Sweden)

Elkin Libardo Ríos

2008-01-01

Full Text Available Entre mayo del 2003 y abril del 2004, en la ciénaga El Eneal, municipio de San Onofre-Sucre, se midieron los perfiles de temperatura del agua, oxígeno disuelto, pH, conductividad eléctrica y salinidad a través de un diseño nictemeral. Se encontró que el sistema es un ambiente completamente mezclado desde el punto de vista térmico debido a la acción de los vientos, de su morfología y de su ubicación cerca de la línea costera. También, se halló que esta ciénaga costera es un ambiente oligohalino en época seca; sin embargo, la mayor parte del tiempo el sistema puede considerarse como un ambiente limnético. En épocas prolongadas de sequía, la salinidad alcanzó su valor máximo de 3,4 ppm, lo cual podría constituir un factor limitante para comunidades de organismos estrictamente limnéticos.

Technical evaluation of the susceptibility of safety-related systems to flooding caused by the failure of non-category 1 systems for the San Onofre Nuclear Power Plant, Unit 1

International Nuclear Information System (INIS)

Latorre, V.R.; Victor, R.A.

1980-11-01

This report documents the technical evaluation of Southern California Edison Company's San Onofre Nuclear Power Plant, Unit 1, to determine whether the failure of any non-Category 1 (seismic) equipment could result in a condition, such as flooding, that might potentially adversely affect the performance of safety-related equipment required for the safe shutdown of the facility or to mitigate the consequences of an accident. Criteria developed by the US Nuclear Regulatory Commission were used to evaluate the acceptability of the existing protection as well as measures taken by Southern California Edison Company to minimize the danger of flooding and to protect safety-related equipment

Systematic evaluation program, status summary report

International Nuclear Information System (INIS)

1983-01-01

Status reports are presented on the systematic evaluation program for the Big Rock Point reactor, Dresden-1 reactor, Dresden-2 reactor, Ginna-1 reactor, Connecticut Yankee reactor, LACBWR reactor, Millstone-1 reactor, Oyster Creek-1 reactor, Palisades-1 reactor, San Onofre-1 reactor, and Rowe Yankee reactor

Evaluation of the integrity of SEP reactor vessels

International Nuclear Information System (INIS)

Hoge, K.G.

1979-12-01

A documented review is presented of the integrity of the 11 reactor pressure vessels covered in the Systematic Evaluation Program. This review deals primarily with the design specifications and quality assurance programs used in the vessel construction and the status of material surveillance programs, pressure-temperature operating limits, and inservice inspection programs of the applicable plants. Several generic items such as PWR overpressurization protection and BWR nozzle and safe-end cracking also are evaluated. The 11 vessels evaluated include Dresden Units 1 and 2, Big Rock Point, Haddam Neck, Yankee Rowe, Oyster Creek, San Onofre 1, LaCrosse, Ginna, Millstone 1, and Palisades

Safety evaluation report related to the full-term operating license for San Onofre Nuclear Generating Station, Unit 1 (Docket No. 50-206)

International Nuclear Information System (INIS)

1991-07-01

The safety evaluation report for the full-term operating license application filed by the Southern California Edison Company and the San Diego Gas and Electric Company has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in San Diego County, California. The staff has evaluated the issues related to the conversion of the provisional operating license to a full-term operating license and concluded that the facility can continue to be operated without endangering the health and safety of the public following the license conversion. 43 refs., 3 figs., 3 tabs

Ten-year rollover of San Onofre inservice testing program for pumps and valves to OM-6 and OM-10

International Nuclear Information System (INIS)

Croy, P.A.; Fischetti, S.; Chiang, D.; Schofield, P.; Barney, D.

1994-01-01

The Pump and Valve Inservice Testing (IST) Program Sat San Onofre, Units 2 and 3, was updated for the second 120-month interval from August 1993 to April 1994. The U.S. Nuclear Regulatory Commission (USNRC) approved the OM-6 and OM-10 Codes in mid-1992. The project for the rollover to these new Codes included several elements: (a) a review of the differences between IWV/IWP and OM-6/OM-10, (b) a comprehensive audit of the IST Program scope for valves, (c) creation of the program and supporting basis documents, the Relief Requests, and implementing procedures, (d) interdivisional coordination, (e) submittal to the USNRC, and (f) training. Subsections IWV and IWP have been used and essentially unchanged for over a decade. The new Code (Parts 1, 6, and 10 called OM-1, OM-6, and OM-10) includes several significant changes from the old Code. Our group identified these differences and drafted revised and reorganized Inservice Testing (IST) Program documents. We also considered USNRC Generic Letter 89-04 (GL 89-04), open-quotes Guidance on Developing Acceptable Inservice Testing Programsclose quotes, and NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, while revising the program. There were six pump relief requires and 13 valve relief requests in the program for the first 10-year interval. For the revised program we needed only one pump relief request (and no valve relief requests). Converting to the 1989 edition of the ASME Code did not require changes to the technical specifications. We revised our Updated Final Safety Analysis Report (UFSAR) to reflect the IST Program for the second 10-year interval. UFSAR changes were minor, consisting of updated references to the Code edition and 10 CFR 50.55a(f), open-quotes Inservice Testing Requirementsclose quotes

SANS facility at the Pitesti 14 MW Triga reactor

International Nuclear Information System (INIS)

Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A.; Grabcev, B.; Todireanu, S.; Constantin, F.; Shvetsov, V.; Popescu, G.

2006-01-01

Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom ≤ λ ≤ 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Q min = 0.005 Angstrom -1 while the maximal value is Q max = 0.5 Angstrom -1 . The relative error is ΔQ/Q min = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation

Cost/benefit analysis of adding a feed-and-bleed capability to Combustion Engineering pressurized-water reactors

International Nuclear Information System (INIS)

Gallup, D.R.; Gahan, E.; Cherdack, R.; Skala, G.

1983-08-01

This report presents the results of a cost/benefit analysis for the addition of a feed-and-bleed capability to the San Onofre Nuclear Generating Station, Unit 2, (SONGS 2). Two cases of feed-and-bleed capability were investigated: 1) adding power-operated relief valves (PORVs) to the pressurizer for depressurization and using the present high-pressure safety-injection (HPSI) system for reactor-coolant-system (RCS) inventory make-up and 2) adding an independent single-train feed-and-bleed system. For the first case, it is estimated that the core-melt frequency would be incrementally reduced by 4.0E-6 per year, a factor of 1.3, at a cost of $2.5 M to $4.3 M depending on when the equipment is installed. For the second case, it is estimated that the core-melt frequency would be incrementally reduced by 1.2E-5 per year, a factor of 3, at a cost of $7.0 M to $10.3 M

Properties of nuclei and elementary particles at low and intermediate energies. Progress report, July 1992--August 1993

International Nuclear Information System (INIS)

Boehm, F.

1993-01-01

Work reported relate to: a 12 ton low energy neutrino detector for neutrino oscillation studies at the San Onofre Reactor Station; new limits on the 17 keV neutrino; time reversal and parity tests for hindered nuclear gamma transitions; and theory of nuclear structure and its application

SCE, PG ampersand E face off with California PUC on shutdown

International Nuclear Information System (INIS)

Anon.

1994-01-01

The California Public Utilities Commission (CPUC) continues its consideration of a proposal to close permanently the San Onofre nuclear station, near San Clemente, California. In its report to the full CPUC, the Division of Ratepayer Advocates (DRA) concluded that continuing to operate San Onofre was not cost-effective compared with the cost of replacement power. The DRA claims the state could save money by closing the plant and utilizing demand-side management programs and power purchases from other utilities to replace the power from San Onofre at a lower cost

Optimized PWR power ascension reload testing

International Nuclear Information System (INIS)

Emery, S.P.; Long, S.W.; Nazareth, V.F.; Herschthal, M.A.

1987-01-01

Reduction in critical path testing time following refueling is actively supported by utilities to increase plant capacity factor and to minimize replacement power costs. Combustion Engineering (C-E) has developed a fast power ascension program (FPAP), which reduces this critical path testing by minimizing holds at intermediate power levels and by automating data acquisition and analysis. A very successful demonstration of the FPAP was performed recently during the cycle 3 startup of Southern California Edison's San Onofre Unit 2 reactor, which resulted in a critical path time savings of ∼ 3 days

The course of a true outage never ran so smooth

International Nuclear Information System (INIS)

Harberts, Craig

1994-01-01

In order to improve the performance of outages at San Onofre Nuclear Generating Station in California, the working structure of the entire organisation has had to be radically altered, in order to bring San Onofre up to standard with other nuclear plants known to be performing well. Working systems were simplified and efficiency improved. Personnel needed to be remotivated to work cooperatively and the outage budget process was revised to include input from all relevant organizations, historical costs and benchmark information from other plants that performed well. Finally, the decision making and teamwork culture has altered radically at San Onofre over the last decade. (UK)

Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

International Nuclear Information System (INIS)

Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

2010-01-01

Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

Managing engineering to meet construction requirements

International Nuclear Information System (INIS)

Martin, D.F.; Houchen, J.D.

1976-01-01

The San Onofre Units 2 and 3 Project schedule is compared with Bechtel's Generic Nuclear Power Plant schedule. This comparison shows that the major delays experienced on the San Onofre Project have resulted from the regulatory process. To date, Engineering has met Construction's requirements and the Project has not experienced any Engineering related delays. The San Onofre Project has been faced with many uncertainties, such as limited site area, high seismic design criteria, new and changing Federal and State regulations, shifts in supplier market conditions and unpredictable supplier performance. Each of these uncertainties has impacted the Engineering effort and jeopardized project schedule goals. The SCE-Bechtel Engineering Management team has acted to mitigate the impact of these uncertainties through use of a cost trend program, simplification of SCE-Bechtel interfaces, close Engineering-Construction coordination, the use of task forces to handle critical supplier problems and the use of additional Engineering personnel, etc. so that Construction requirements have been met

Comparison of SANS instruments at reactors and pulsed sources

International Nuclear Information System (INIS)

Thiyagarajan, P.; Epperson, J.E.; Crawford, R.K.; Carpenter, J.M.; Hjelm, R.P. Jr.

1992-01-01

Small angle neutron scattering is a general purpose technique to study long range fluctuations and hence has been applied in almost every field of science for material characterization. SANS instruments can be built at steady state reactors and at the pulsed neutron sources where time-of-flight (TOF) techniques are used. The steady state instruments usually give data over small q ranges and in order to cover a large q range these instruments have to be reconfigured several times and SANS measurements have to be made. These instruments have provided better resolution and higher data rates within their restricted q ranges until now, but the TOF instruments are now developing to comparable performance. The TOF-SANS instruments, by using a wide band of wavelengths, can cover a wide dynamic q range in a single measurement. This is a big advantage for studying systems that are changing and those which cannot be exactly reproduced. This paper compares the design concepts and performances of these two types of instruments

Simulation of SONGS unit 2/3 NSSS with RETACT

International Nuclear Information System (INIS)

Fakory, M.R.; Olmos, J.

1991-01-01

RETACT Code which is a major code for real time simulation of thermal-hydraulic phenomena has been enhanced and configured for the first time for Simulation of the Nuclear Steam Supply System (NSSS) of C-E designed PWRs at San Onofre Nuclear Generating Station. SONGS Unit 2/3 Simulator was upgraded for thermal-hydraulic and containment models as well as the instructor station. In this paper the simulator results for various transients and accidents were benchmarked against plant data, the comparison for some of the benchmarkings including steam generator level swell/shrink, and loss-of-coolant accident are presented

A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

International Nuclear Information System (INIS)

Basha, H.S.; Manahan, M.P.

1992-01-01

In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

"Life without nuclear power": A nuclear plant retirement formulation model and guide based on economics. San Onofre Nuclear Generating Station case: Economic impacts and reliability considerations leading to plant retirement

Science.gov (United States)

Wasko, Frank

Traditionally, electric utilities have been slow to change and very bureaucratic in nature. This culture, in and of itself, has now contributed to a high percentage of United States electric utilities operating uneconomical nuclear plants (Crooks, 2014). The economic picture behind owning and operating United States nuclear plants is less than favorable for many reasons including rising fuel, capital and operating costs (EUCG, 2012). This doctoral dissertation is specifically focused on life without nuclear power. The purpose of this dissertation is to create a model and guide that will provide electric utilities who currently operate or will operate uneconomical nuclear plants the opportunity to economically assess whether or not their nuclear plant should be retired. This economic assessment and stakeholder analysis will provide local government, academia and communities the opportunity to understand how Southern California Edison (SCE) embraced system upgrade import and "voltage support" opportunities to replace "base load" generation from San Onofre Nuclear Generating Station (SONGS) versus building new replacement generation facilities. This model and guide will help eliminate the need to build large replacement generation units as demonstrated in the SONGS case analysis. The application of The Nuclear Power Retirement Model and Guide will provide electric utilities with economic assessment parameters and an evaluation assessment progression needed to better evaluate when an uneconomical nuclear plant should be retired. It will provide electric utilities the opportunity to utilize sound policy, planning and development skill sets when making this difficult decision. There are currently 62 nuclear power plants (with 100 nuclear reactors) operating in the United States (EIA, 2014). From this group, 38 are at risk of early retirement based on the work of Cooper (2013). As demonstrated in my model, 35 of the 38 nuclear power plants qualify to move to the economic

A review on research activities using the SANS spectrometer in transmission geometry at ET-RR-1 reactor

International Nuclear Information System (INIS)

Adib, M.

1999-01-01

The phased double rotor facility operating at ET-RR-1 reactor (2MW) was rearranged to operate as SANS spectrometer in transmission geometry. The rotors are suspended in magnetic fields and are spinning up to 16,000 rpm producing bursts of polyenergetic neutrons with wavelengths from 0.2 nm to 6.5 nm and beam divergence of 17' on the sample. The review on research activities using the SANS spectrometer and its applications for powder particle size determination and the long wavelength fluctuation of magnetization of the Fe-Ni alloys are discussed. (author)

Recent developments: Industry briefs

International Nuclear Information System (INIS)

Anon.

1992-01-01

This is the February 1992 'Industry Briefs' portion of the 'Recent Developments' section. Issues mentioned are: (1) closure of San Onofre Unit 1, (2) start-up of Penly Unit 2, (3) signing of a safeguards agreement with North Korea, (4) Canadian nuclear activities in Romania, and (5) the merger of two Japanese fuel cycle companies

Importance of deposit information in the design and execution of steam generator chemical cleaning

International Nuclear Information System (INIS)

Flores, O.; Remark, J.

1997-01-01

During the planning stages of the chemical cleaning of the San Onofre Nuclear Generating Station (SONGS) units 2 and 3 steam generators, it was determined that an understanding of the steam generator deposit loading and composition was essential to the design and success of the project. It was also determined that qualification testing, preferably with actual deposits from the SONGS steam generators, was also essential. SONGS units 2 and 3 have Combustion Engineering (CE)-designed pressurized water reactors. Each unit has two CE model 3410 steam generators. Each steam generator has 9350 alloy 600 tubes with 1.9-cm (3/4 in.) outside diameter. Unit 2 began commercial operation in 1983, and unit 3, in 1984. The purpose of this technical paper is to explain the effort and methodology for deposit composition, characterization, and quantification. In addition, the deposit qualification testing and design of the cleaning are discussed

International training course on implementation of state systems of accounting for and control of nuclear materials: proceedings

International Nuclear Information System (INIS)

1986-06-01

This report incorporates all lectures and presentations at the International Training Course on Implementation of State Systems of Accounting for and Control of Nuclear Materials held June 3 through June 21, 1985, at Santa Fe and Los Alamos, New Mexico, and San Clemente, California. Authorized by the US Nuclear Non-Proliferation Act and sponsored by the US Department of Energy in cooperation with the International Atomic Energy Agency, the Course was developed to provide practical training in the design, implementation, and operation of a state system of nuclear materials accountability and control that satisfies both national and international safeguards requirements. Major emphasis for the 1985 course was placed on safeguards methods used at item-control facilities, particularly nuclear power generating stations and test reactors. An introduction to safeguards methods used at bulk handling facilities, particularly low-enriched uranium conversion and fuel fabrication plants, was also included. The course was conducted by the University of California's Los Alamos National Laboratory and the Southern California Edison Company. Tours and demonstrations were arranged at the Los Alamos National Laboratory, Los Alamos, New Mexico, and the San Onofre Nuclear Generating Station, San Clemente, California

International training course on implementation of state systems of accounting for and control of nuclear materials: proceedings

Energy Technology Data Exchange (ETDEWEB)

1986-06-01

This report incorporates all lectures and presentations at the International Training Course on Implementation of State Systems of Accounting for and Control of Nuclear Materials held June 3 through June 21, 1985, at Santa Fe and Los Alamos, New Mexico, and San Clemente, California. Authorized by the US Nuclear Non-Proliferation Act and sponsored by the US Department of Energy in cooperation with the International Atomic Energy Agency, the Course was developed to provide practical training in the design, implementation, and operation of a state system of nuclear materials accountability and control that satisfies both national and international safeguards requirements. Major emphasis for the 1985 course was placed on safeguards methods used at item-control facilities, particularly nuclear power generating stations and test reactors. An introduction to safeguards methods used at bulk handling facilities, particularly low-enriched uranium conversion and fuel fabrication plants, was also included. The course was conducted by the University of California's Los Alamos National Laboratory and the Southern California Edison Company. Tours and demonstrations were arranged at the Los Alamos National Laboratory, Los Alamos, New Mexico, and the San Onofre Nuclear Generating Station, San Clemente, California.

Towards Compact Antineutrino Detectors for Safeguarding Nuclear Reactors

International Nuclear Information System (INIS)

Meijer, R.J. de; Smit, F.D.; Woertche, H.J.

2010-01-01

In 2008 the IAEA Division of Technical Support convened a Workshop on Antineutrino Detection for Safeguards Applications. Two of the recommendations expressed that IAEA should consider antineutrino detection and monitoring in its current R and D program for safeguarding bulk-process reactors, and consider antineutrino detection and monitoring in its Safeguards by Design approaches for power and fissile inventory monitoring of new and next generation reactors. The workshop came to these recommendations after having assessed the results obtained at the San Onofre Nuclear Generator Station (SONGS) in California. A 600 litre, 10% efficiency detector, placed at 25m from the core was shown to record 300 net antineutrino events per day. The 2*2.5*2.5 m 3 footprint of the detector and the required below background operation, prevents an easy deployment at reactors. Moreover it does not provide spatial information of the fissile inventory and, because of the shape of a PBMR reactor, would not be representative for such type of reactor. A solution to this drawback is to develop more efficient detectors that are less bulky and less sensitive to cosmic and natural radiation backgrounds. Antineutrino detection in the SONGS detector is based on the capture of antineutrinos by a proton resulting in a positron and neutron. In the SONGS detector the positron and neutron are detected by secondary gamma-rays. The efficiency of the SONGS detector is largely dominated by the low efficiency for gamma detection high background sensitivity We are investigating two methods to resolve this problem, both leading to more compact detectors, which in a modular set up also will provide spatial information. One is based on detecting the positrons on their slowdown signal and the neutrons by capturing in 10 B or 6 Li, resulting in alpha-emission. The drawback for standard liquid scintillators doped with e.g. B is the low flame point of the solvent and the strong quenching of the alpha signal. Our

Containment at the Source during Waste Volume Reduction of Large Radioactive Components Using Oxylance High-Temperature Cutting Equipment - 13595

International Nuclear Information System (INIS)

Keeney, G. Neil

2013-01-01

As a waste-volume reduction and management technique, highly contaminated Control Element Drive Mechanism (CEDM) housings were severed from the Reactor Pressure Vessel Head (RPVH) inside the San Onofre Unit 2 primary containment utilizing Oxylance high-temperature cutting equipment and techniques. Presented are relevant data concerning: - Radiological profiles of the RPVH and individual CEDMs; - Design overviews of the engineering controls and the specialized confinement housings; - Utilization of specialized shielding; - Observations of apparent metallurgical-contamination coalescence phenomena at high temperatures resulting in positive control over loose-surface contamination conditions; - General results of radiological and industrial hygiene air sampling and monitoring; - Collective dose and personnel contamination event statistics; - Lessons learned. (author)

Containment at the Source during Waste Volume Reduction of Large Radioactive Components Using Oxylance High-Temperature Cutting Equipment - 13595

Energy Technology Data Exchange (ETDEWEB)

Keeney, G. Neil [Health Physicist, HazMat CATS, LLC (United States)

2013-07-01

As a waste-volume reduction and management technique, highly contaminated Control Element Drive Mechanism (CEDM) housings were severed from the Reactor Pressure Vessel Head (RPVH) inside the San Onofre Unit 2 primary containment utilizing Oxylance high-temperature cutting equipment and techniques. Presented are relevant data concerning: - Radiological profiles of the RPVH and individual CEDMs; - Design overviews of the engineering controls and the specialized confinement housings; - Utilization of specialized shielding; - Observations of apparent metallurgical-contamination coalescence phenomena at high temperatures resulting in positive control over loose-surface contamination conditions; - General results of radiological and industrial hygiene air sampling and monitoring; - Collective dose and personnel contamination event statistics; - Lessons learned. (author)

Implementation of an RHR/LPSI pump coupling retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; Koch, R.P.; Orewyler, R.; Tipton, J.W.

1994-01-01

Nuclear plant operating experience has shown the RHR and LPSI services to be very demanding on pumps. The systems handle borated water at high temperatures and pressures with frequent step changes in both temperature and pressure. Additionally, the industry trend towards reduced flow rates during plant mid-loop (reduced inventory) conditions has resulted in extended pump operation at flow rates significantly below the pump best efficiency point flow. Operation at these low flow fates is known to cause high thrust loads and large shaft deflections. The combination of these and other factors have resulted in short mechanical seal life and short motor bearing life, thus requiring frequent pump and motor maintenance. For many nuclear plants, including Southern California Edison's (SCE) San Onofre Units 2 and 3, these pumps have represented a major operations and maintenance (O ampersand M) expenditure and a significant source of radiation exposure to plant personnel. SCE management determined that a pump upgrade was justified to reduce the O ampersand M costs and to improve plant availability. SCE decided to proceed with a pump retrofit program to improve the pump maintainability, reliability and availability. Installation was completed for four LPSI pumps at San Onofre Units 2 and 3 during the Cycle 7 refueling outages in 1993. A key to the program's success was the removal of many traditional supplier and customer barriers and revision of supplier and customer roles to create a unified team. This paper traces the RHR/LPSI retrofit program for San Onofre from problem identification to project implementation. The team approach used for this program and the lessons learned may be useful to other utilities and vendors when evaluating or implementing system and equipment upgrades

Risk of nuclear power generation as business (continued)

International Nuclear Information System (INIS)

Sato, Satoshi

2017-01-01

This paper described the following: (1) fleet formation of power companies that operate nuclear power plants in the U.S., (2) collaboration, competition, and merger between plant makers, (3) stress corrosion cracking of stream generators for PWR and their thin heat transfer tubes, especially stress corrosion cracking under primary cooling water environment (PWSCC), and (4) replacement project from Inconel 600 MA to Inconel 600 TT or 690 TT of steam generator thin heat transfer tubes of PWR plants in the U.S. and others. In addition, it described the troubles at San Onofre Nuclear Power Station in California: wear of steam generator thin tubes of Units 2 and 3, and leakage from primary system to secondary system of Unit 3, and permanent shutdown. It also described the detail of damages compensation talks between South California Edison Company that operates San Onofre nuclear power plant and Mitsubishi Heavy Industries Ltd. which supplied the steam generator. Although the operation of the 1.7 million kW plant became impossible due to the bud shedding of nuclear power renaissance, these troubles might have saved the nightmare of drifting on the way. (A.O.)

Advances in seismic criteria to qualify structures, systems and components in operating reactors

International Nuclear Information System (INIS)

Manrique, M.A.; Bak, W.R.

1989-01-01

This paper describes improved seismic evaluation criteria and analysis methodologies used as part of the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1. The plant had originally been designed for 0.25 g ground acceleration and was required to be upgraded to a 0.67 g ground acceleration as part of the plant's Long Term Service Seismic Reevaluation Program. The application of the criteria and methods described in this paper to demonstrate the seismic capability of the plant resulted in efficient plant modifications with considerable cost savings to the plant owner. The NRC accepted these criteria and methods based on favorable results of reviews, audits and independent verification of the theories, bases and implementation procedures of the proposed criteria and analysis methods

Seismic safety margins research program. Project I SONGS 1 AFWS Project

International Nuclear Information System (INIS)

Chuang, T.Y.; Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Cummings, G.E.; Wells, J.E.

1981-01-01

The seismic qualification requirements of auxiliary feedwater systems (AFWS) of Pressurized Water Reactors (PWR) were developed over a number of years. These are formalized in the publication General Design Criteria (Appendix A to 10CFR50). The full recognition of the system as an engineered safety feature did not occur until publication of the Standard Review Plan (1975). Efforts to determine how to backfit seismic requirements to earlier plants has been undertaken primarily in the Systematic Evaluation Program (SEP) for a limited number of operating reactors. Nuclear Reactor Research (RES) and NRR have requested LLNL to perform a probabilistic study on the AFWS of San Onofre Nuclear Generating Station (SONGS) Unit 1 utilizing the tools developed by the Seismic Safety Margins Research Program (SSMRP). The main objectives of this project are to: identify the weak links of AFWS; compare the failure probabilities of SONGS 1 and Zion 1 AFWS: and compare the seismic responses due to different input spectra and design values

Southern California Edison instrument setpoint program

International Nuclear Information System (INIS)

Bockhorst, R.M.; Quinn, E.L.

1991-01-01

In November of 1989, the US Nuclear Regulatory Commission (NRC) conducted an electrical safety system functional inspection (ESSFI) at the San Onofre nuclear generating station (SONGS), which was followed by an NRC audit on instrument setpoint methodology in January 1991. Units 2 and 3 at SONGS are 1100-MW(electric) Combustion Engineering (C-E) pressurized water reactors (PWRs) operated by Southern California Edison (SCE). The purpose of this paper is to summarize the results of the NRC audit and SCE's follow-up activities. The NRC team inspection reinforced the need to address several areas relative to the SCE setpoint program. The calculations withstood the intensive examination of four NRC inspectors for 2 weeks and only a few minor editorial-type problems were noted. Not one of the calculated plant protections system setpoints will change as a result of the audit. There were no questions raised relative to setpoint methodology

Experiences with drug testing at a nuclear power plant

International Nuclear Information System (INIS)

Ray, H.B.

1987-01-01

After more than 2 yr of operation of a drug testing program at the San Onofre nuclear power plant site, the Southern California Edison Co. has had a number of experiences and lessons considered valuable. The drug testing program at San Onofre, implemented in September of 1984, continues in essentially the same form today. Prior to describing the program, the paper reviews several underlying issues that believed to be simultaneously satisfied by the program: trustworthiness, fitness and safety, public trust, and privacy and search. The overall drug testing program, periodic drug monitoring program, and unannounced drug testing program are described. In addition to the obvious features of a good drug testing program, which are described in the EEI guide, it is essential to consider such issues as the stated program rationale, employee relations, and disciplinary action measures when contemplating or engaging in drug testing at nuclear power plants

Use of the Safety Monitor in operational decision-making at a nuclear generating facility

International Nuclear Information System (INIS)

Chien, Shan H.; Hook, Thomas G.; Lee, Roger J.

1998-01-01

The utilization of Safety Monitor at a nuclear generating facility in 1994 revolutionized the way US nuclear power plants manage configuration risks. At Southern California Edison (SCE) Company's San Onofre Nuclear Generating Station, it transformed probabilistic risk assessment (PRA) from a retrospective tool for understanding past risk into a prospective tool for controlling future risk. Since that time, many other nuclear utilities have taken aggressive steps in using PRA better to understand and manage risks associated with plant operation and maintenance. These utilities have employed a variety of methods ranging from systems similar to San Onofre's Safety Monitor to systems dramatically different in both technology and philosophy. In the development and use of its Safety Monitor, SCE has been guided by two philosophical goals: (1) maximize the objectivity of PRA-informed decision-making relative to managing configuration risks, and (2) ensure that risks are managed conservatively

PREFACE: SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor: Devoted to the 75th anniversary of Yu M Ostanevich's birth

Science.gov (United States)

Gordely, Valentin; Kuklin, Alexander; Balasoiu, Maria

2012-03-01

The Second International Workshop 'SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor', devoted to the 75th anniversary of the birth of Professor Yu M Ostanevich (1936-1992), an outstanding neutron physicist and the founder of small-angle neutron scattering (field, group, and instrument) at JINR FLNPh, was held on 27-30 May at the Frank Laboratory of Neutron Physics. The first Workshop was held in October 2006. Research groups from different neutron centers, universities and research institutes across Europe presented more than 35 oral and poster presentations describing scientific and methodological results. Most of them were obtained with the help of the YuMO instrument before the IBR-2 shutdown in 2006. For the last four years the IBR-2 reactor has been shut down for refurbishment. At the end of 2010 the physical launch of the IBR-2M reactor was finally realized. Nowadays the small-angle neutron scattering (SANS) technique is applied to a wide range of scientific problems in condensed matter, soft condensed matter, biology and nanotechnology, and despite the fact that there are currently over 30 SANS instruments in operation worldwide at both reactor and spallation sources, the demand for beam-time is considerably higher than the time available. It must be remembered, however, that as the first SANS machine on a steady-state reactor was constructed at the Institute Laue Langevin, Grenoble, the first SANS instrument on a 'white' neutron pulsed beam was accomplished at the Joint Institute for Nuclear Research at the IBR-30 reactor, beamline N5. During the meeting Yu M Ostanevich's determinative and crucial contribution to the construction of spectrometers at the IBR-2 high-pulsed reactor was presented, as well as his contribution to the development of the time-of-flight (TOF) small-angle scattering technique, and a selection of other scientific areas. His leadership and outstanding scientific achievements in applications of the

Nuclear Regulatory Commission issuances. Volume 17, No. 3

International Nuclear Information System (INIS)

1983-03-01

This report contains the Issuances received during March 1983 from the Commission (CLI), the Atomic Safety and Licensing Appeal Boards (ALAB), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judge (ALJ), the Directors Decisions (DD), and the Denials of Petition for Rulemaking (DPRM). The Issuances concerned the following facilities: Three Mile Island Nuclear Station, Unit No. 1; Comanche Peak Steam Electric Station, Units 1 and 2; Vallecitos Nuclear Center; Floating Nuclear Power Plants; San Onofre Nuclear Generating Station, Units 2 and 3; Point Beach Nuclear Plant, Unit 1; Perry Nuclear Power Plant, Units 1 and 2; Shoreham Nuclear Power Station, Unit 1; Western New York Nuclear Service Center; Limerick Generating Station, Units 1 and 2; Seabrook Station, Units 1 and 2; Black Fox Station, Units 1 and 2; WmH Zimmer Nuclear Power Station, Unit 1; WPPSS Nuclear Project No. 1; Zion Nuclear Plant, Units 1 and 2; and South Texas Project, Units 1 and 2

75 FR 69136 - Southern California Edison Company, San Onofre Nuclear Generating Station, Units 2 and 3...

Science.gov (United States)

2010-11-10

... radiation exposures to plant workers and members of the public. Therefore, no radiological impacts are..., socioeconomic conditions, and minority- and low-income populations in the vicinity of SONGS 2 and 3 would also...

San Diego, California 1/3 arc-second DEM

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The 1/3-second San Diego, California Elevation Grid provides bathymetric data in ASCII raster format of 1/3-second resolution in geographic coordinates. This grid is...

Gas-cooled reactor coolant circulator and blower technology. Proceedings of a specialists meeting held in San Diego 30 November - 2 December 1987

Energy Technology Data Exchange (ETDEWEB)

1988-08-01

In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs.

Overview of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

2016-01-01

The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

Innovative probabilistic risk assessment applications: barrier impairments and fracture toughness. 2. Demolition Debris and Tornado Missile Hazard During Decommissioning

International Nuclear Information System (INIS)

Calhoun, David; Shepherd, Stephen

2001-01-01

During their operating lives, nuclear power plants typically maintain a high level of control over the amount of debris that is allowed to accumulate at the plant site. Although primarily intended to reduce the potential for fire damage, some plants also rely on these controls to limit the damage that could be caused during a tornado from missiles generated from loose debris. Demolition work associated with power plant decommissioning inevitably increases the quantity of debris. When bulk commodities such as piping and electrical distribution components are demolished, they are subject to various staging, handling, and storage processes before they can be released from the site. The demolition of plant structures dramatically increases the quantity of loose steel and concrete debris. For the foreseeable future, all plants that undertake decommissioning will have spent-fuel assemblies present on the plant site during the demolition project whether the spent fuel remains stored in a spent-fuel pool or is transferred to an independent spent-fuel storage installation (ISFSI). Under present regulations, protection from tornado missiles would be required for both types of spent-fuel storage. In addition, a small proportion of decommissioning plants will have operating units in close proximity. Licensing commitments for tornado missile protection may mandate controls on the production or storage of demolition debris. This paper presents a case study of the San Onofre Nuclear Generating Station (Fig. 1). Tornado missile protection licensing commitments from three types of facilities will be in force during the decommissioning of San Onofre Unit 1 (Unit 1): 1. Unit 1, under a possession-only license; 2. an ISFSI that will eventually store spent fuel from Unit 1; 3. San Onofre Operating Unit 2 (Unit 2) and San Onofre Operating Unit 3 (Unit 3). Together, these three facilities illustrate the range of impacts that licensing commitments designed for tornado protection may

Lessons learned from the seismic reevaluation of San Onofre Nuclear Generating Station, Unit 1

International Nuclear Information System (INIS)

Russell, M.J.; Shieh, L.C.; Tsai, N.C.; Cheng, T.M.

1987-01-01

SCE developed site specific ground response spectra based on the Housner response spectra. These were modified as a result of NRC review to account for uncertainties in a limited frequency range, and were applied in a conventional manner to the development of time histories for structural analysis. Frequency-dependent soil structure interaction analyses were performed for the reactor building. Frequency-independent soil stiffness and damping values were used in conventional analyses of the turbine building. Conventional acceptance criteria were applied to concrete structures. Supplemental acceptance criteria were developed for the steel structures; these criteria allowed a limited amount of ductility in isolated members and were applied only if the connections were maintained in the elastic range. Consideration of the effect of the ductility on the total structural response was required. Application of these criteria and methods required only a moderate level of reinforcement of steel structures in order to assure adequate structural integrity. The development, review and approval of supplemental criteria and methods was extented to the reevaluation of piping and equipment, with an emphasis on piping. (orig./HP)

Flux and fluence determination using the material scrapings approach

International Nuclear Information System (INIS)

Basha, H.S.; Manahan, M.P.

1992-01-01

The conventional approach to flux determination is to use high-purity dosimeters to characterize the neutron field. This paper presents an alternative approach called the scraping method. This method consists of taking scraping samples from an in-service component and using this material to measure the specific activity for various reactions. This approach enables the determination of the neutron flux and fluence incident on any component for which small chips of material can be safely obtained. It offers a capability for determining the neutron flux for components such as reactor internals without destructively removing them from service. The scrapings methodology was benchmarked by comparison with the results obtained using conventional dosimetry data from the San Onofre nuclear generation station Unit 2 (SONGS-2). Additionally, since the goal of any reactor physics analysis is to reduce uncertainty to the extent practical, it is important that the best available cross-section library be used. The fast flux calculated-to-experimental (C/E) ratios at the SONGS-297-deg in-vessel surveillance capsule and the REACTOR-X 90-deg ex-vessel dosimetry positions were studied for several cross-section libraries, including BIGLE-80, SAILOR, and ELXSIR. REACTOR-X is a pressurized water reactor power plant currently operating in the US

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Estimation of immersion dose to the Tin mining in the seashore of Bangka island from NPP operation

International Nuclear Information System (INIS)

Nurokhim; Erwansyah Lubis

2015-01-01

The estimation of immersion dose to the Tin (Sn) mining in the seashore of Bangka island using concentrations factor methods was carried out. In the estimation, the source-term of effluent released to Pacific ocean from Diablo Canyon and San Onofre Nuclear Power Plant (NPP) operations was used. The results indicated that the Sn mining with immersion in the sea water of 2, 4 and 6 hours per day will receives maximum effective dose of 4.45 × 10 -3 %; 8.90 ×10 -3 % and 1.34 × 10 -2 % from dose constraint of 0.3 mSv per years. The probability of cancer happen for individual are 2.67 × 10 -8 ; 5.34 × 10 -8 and 8.01 × 10 -8 respectively if the working hours are 2, 4 dan 6 hours per days as long as 40 years (reactor lifetime) . These data give the information if there are 50 millions of Sn mining, the potential of miner will receive a fatal cancer is around 4 persons. According to the population of Sn mining is very small, so the probabilities of fatal cancer is unsignificant. (author)

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

High intensity multi beam design of SANS instrument for Dhruva reactor

Energy Technology Data Exchange (ETDEWEB)

Abbas, Sohrab, E-mail: abbas@barc.gov.in; Aswal, V. K. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Désert, S. [Laboratoire Leon Brillouin, CEA, Saclay, 91191 (France)

2016-05-23

A new and versatile design of Small Angle Neutron Scattering (SANS) instrument based on utilization of multi-beam is presented. The multi-pinholes and multi-slits as SANS collimator for medium flux Dhruva rearctor have been proposed and their designs have been validated using McStas simulations. Various instrument configurations to achieve different minimum wave vector transfers in scattering experiments are envisioned. These options enable smooth access to minimum wave vector transfers as low as ~ 6×10{sup −4} Å{sup −1} with a significant improvement in neutron intensity, allowing faster measurements. Such angularly well defined and intense neutron beam will allow faster SANS studies of agglomerates larger than few tens of nm.

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

2013-09-30

This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

International Nuclear Information System (INIS)

Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z.

1999-01-01

The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

Predicting the severity of nuclear power plant transients using nearest neighbors modeling optimized by genetic algorithms on a parallel computer

International Nuclear Information System (INIS)

Lin, J.; Bartal, Y.; Uhrig, R.E.

1995-01-01

The importance of automatic diagnostic systems for nuclear power plants (NPPs) has been discussed in numerous studies, and various such systems have been proposed. None of those systems were designed to predict the severity of the diagnosed scenario. A classification and severity prediction system for NPP transients is developed. The system is based on nearest neighbors modeling, which is optimized using genetic algorithms. The optimization process is used to determine the most important variables for each of the transient types analyzed. An enhanced version of the genetic algorithms is used in which a local downhill search is performed to further increase the accuracy achieved. The genetic algorithms search was implemented on a massively parallel supercomputer, the KSR1-64, to perform the analysis in a reasonable time. The data for this study were supplied by the high-fidelity simulator of the San Onofre unit 1 pressurized water reactor

San Francisco folio, California, Tamalpais, San Francisco, Concord, San Mateo, and Haywards quadrangles

Science.gov (United States)

Lawson, Andrew Cowper

1914-01-01

The five sheets of the San Francisco folio the Tamalpais, Ban Francisco, Concord, Ban Mateo, and Haywards sheets map a territory lying between latitude 37° 30' and 38° and longitude 122° and 122° 45'. Large parts of four of these sheets cover the waters of the Bay of San Francisco or of the adjacent Pacific Ocean. (See fig. 1.) Within the area mapped are the cities of San Francisco, Oakland, Berkeley, Alameda, Ban Rafael, and San Mateo, and many smaller towns and villages. These cities, which have a population aggregating about 750,000, together form the largest and most important center of commercial and industrial activity on the west coast of the United States. The natural advantages afforded by a great harbor, where the railways from the east meet the ships from all ports of the world, have determined the site of a flourishing cosmopolitan, commercial city on the shores of San Francisco Bay. The bay is encircled by hilly and mountainous country diversified by fertile valley lands and divides the territory mapped into two rather contrasted parts, the western part being again divided by the Golden Gate. It will therefore be convenient to sketch the geographic features under four headings (1) the area east of San Francisco Bay; (2) the San Francisco Peninsula; (3) the Marin Peninsula; (4) San Francisco Bay. (See fig. 2.)

CONTAIN 2.0 code release and the transition to licensing

International Nuclear Information System (INIS)

Murata, K.K.; Griffith, R.O.; Bergeron, K.D.; Tills, J.

1997-10-01

CONTAIN is a reactor accident simulation code developed by Sandia National Laboratories under US Nuclear Regulatory Commission (USNRC) sponsorship to provide integrated analysis of containment phenomena, including those related to nuclear reactor containment loads and radiological source terms. The recently released CONTAIN 2.0 code version represents a significant advance in CONTAIN modeling capabilities over the last major code release (CONTAIN 1.12V). The new modeling capabilities are discussed here. The principal motivation for many of the recent model improvements has been to allow CONTAIN to model the special features in advanced light water reactor (ALWR) designs. The work done in this area is also summarized. In addition to the ALWR work, the USNRC is currently engaged in an effort to qualify CONTAIN for more general use in licensing, with the intent of supplementing or possibly replacing traditional licensing codes. To qualify the CONTAIN code for licensing applications, studies utilizing CONTAIN 2.0 are in progress. A number of results from this effort are presented in this paper to illustrate the code capabilities. In particular, CONTAIN calculations of the NUPEC M-8-1 and ISP-23 experiments and CVTR test number-sign 3 are presented to illustrate (1) the ability of CONTAIN to model non-uniform gas density and/or temperature distributions, and (2) the relationship between such gas distributions and containment loads. CONTAIN and CONTEMPT predictions for a large break loss of coolant accident scenario in the San Onofre plant are also compared

33 CFR 165.1141 - Safety Zone; San Clemente 3 NM Safety Zone, San Clemente Island, CA.

Science.gov (United States)

2010-07-01

...) Location. The following area is a safety zone: All waters of the Pacific Ocean surrounding San Clemente Island, from surface to bottom, extending from the high tide line on the island seaward 3 NM. The zone... 3 NM from the high tide line to 33°02.82′ N, 118°30.65′ W; thence 33°01.29′ N, 118°33.88′ W; thence...

SANS-1 Experimental reports of 2000

International Nuclear Information System (INIS)

Willumeit, R.; Haramus, V.

2001-01-01

The instrument SANS-1 at the Geesthacht neutron facility GeNF was used for scattering experiments in 2000 at 196 of 200 days of reactor and cold source operation. The utilisation was shared between the in-house R and D program and user groups from different universities and research centers. These measurements were performed and analysed either by guest scientists or GKSS staff. The focus of the work in 2000 at the experiment SANS-1 was the structural investigation of hydrogen containing substances such as biological macromolecules (ribosomes, protein-RNA-complexes, protein solutions, glycolipids and membranes), molecules which are important in the fields of environmental research (refractoric organic substances) and technical chemistry (surfactants, micelles). (orig.) [de

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

Energy Technology Data Exchange (ETDEWEB)

Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

1999-08-01

The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

Monitoring laundry for fleas

International Nuclear Information System (INIS)

Cooper, T.L.; Goldin, E.M.; Warnock, R.V.

1987-01-01

The San Onofre Unit 3 nuclear plant experienced fuel cladding failures during its first fuel cycle. When primary systems were opened for maintenance and refueling, areas on site were contaminated with fleas - tiny, highly radioactive fuel fragments. While highly mobile, these fragments demonstrated an affinity to attach to some surfaces and clothing. This paper discusses the problems encountered in detecting fleas in the presence of residual activation and fission product contamination on laundered protective clothing. A novel approach to overcome those inherent problems is presented and a home-built laundry monitor utilizing off-the-shelf components is described

Compact High Resolution SANS using very cold neutrons (VCN-SANS)

International Nuclear Information System (INIS)

Kennedy, S.; Yamada, M.; Iwashita, Y.; Geltenbort, P.; Bleuel, M.; Shimizu, H.

2011-01-01

SANS (Small Angle Neutron Scattering) is a popular method for elucidation of nano-scale structures. However science continually challenges SANS for higher performance, prompting exploration of ever-more exotic and expensive technologies. We propose a compact high resolution SANS, using very cold neutrons, magnetic focusing lens and a wide-angle spherical detector. This system will compete with modern 40 m pinhole SANS in one tenth of the length, matching minimum Q, Q-resolution and dynamic range. It will also probe dynamics using the MIEZE method. Our prototype lens (a rotating permanent-magnet sextupole), focuses a pulsed neutron beam over 3-5 nm wavelength and has measured SANS from micelles and polymer blends. (authors)

Country report: utilization of MINT's research reactor

International Nuclear Information System (INIS)

Ab Khalic b Hj Wood; Adnan b Bukhari; Wee Boon Siong

2004-01-01

MINT has only one research reactor, i.e. TRIGA MKII reactor, equipped with various neutron irradiation facilities such as rotary rack and rabbit system. Apart from counting facilities for NAA work, other facilities available for the respective studies include facilities for neutron radiography and SANS. At Present most of reactor operation time has been utilized for samples irradiation related to the NAA application. Majority of the samples are from MINT analytical chemistry laboratory where the present authors work, and the rest of the samples are from local universities. They provide analytical chemistry services for other government departments as well as private companies. In order to improve the reactor utilization, the management of MINT has formed Reactor Interest Group (RIG) at the national level in 2002, which embraces members from various institutions in this country. To support the RIG activities, MINT provides seed funding to finance various activities for the reactor utilization, which include financing project to make use of SANS, neutron radiography and radioisotopes production (mainly for tracer studies carried out by MINT's tracer group) facilities, and funding for basic study in BNCT. (author)

Small-angle neutron scattering instrument of Institute for Solid State Physics, the Univeristy of Tokyo (SANS-U) and its application to biology

Energy Technology Data Exchange (ETDEWEB)

Ito, Yuji; Imai, Masayuki; Takahashi, Shiro [Univ. of Tokyo, Tokai Naka Ibaraki (Japan)

1994-12-31

A small-angle neutron spectrometer (SANS-U) suitable for the study of mesoscopic structure in the field of polymer chemistry and biology, has been constructed at the guide hall of JRR-3M reactor at the Japan Atomic Energy Research Institute. The instrument is 32m long and utilizes a mechanical velocity selector and pinhole collimation to provide a continuous beam with variable wavelength in the range from 5 to 10{Angstrom}. The neutron detector is a 65 x 65cm{sup 2} 2D position sensitive proportional counter. The practical Q range of SANS-U is 0.0008 to 0.45{Angstrom}{sup -1}. The design, characteristics and performance of SANS-U are described with some biological studies using SANS-U.

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

2002-01-01

Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

Safety evaluation report related to the operation of San Onofre Nuclear Generating Station, Units 2 and 3. Docket Nos. 50-361 and 50-362, Southern California Edison Company, et al

International Nuclear Information System (INIS)

1982-06-01

Information is presented concerning seismic design of structures, components, and equipment; reactor coolant system; conduct of operations; initial test program; quality assurance; and Three Mile Island-2 requirements

Power plant cooling systems: trends and challenges

International Nuclear Information System (INIS)

Rittenhouse, R.C.

1979-01-01

A novel design for an intake and discharge system at the Belle River plant is described followed by a general discussion of water intake screens and porous dikes for screening fish and zooplankton. The intake system for the San Onofre PWR plant is described and the state regulations controlling the use of water for power plants is discussed. The use of sewage effluent as a source of cooling water is mentioned with reference to the Palo Verde plant. Progress in dry cooling and a new wet/dry tower due to be installed at the San Juan plant towards the end of this year, complete the survey

Neutron beam applications - Polymer study and sample environment development for HANARO SANS instrument

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong Doo [Kyunghee University, Seoul (Korea); Char, Kook Heon [Seoul National University, Seoul (Korea)

2000-04-01

A new SANS instrument will be installed in HANARO reactor near future and in parallel it is necessary to develop the sample environment facilities. One of the basic items is the equipment to control the sample temperature of cell block with auto-sample changer. It is required to develop a control software for this purpose. In addition, softwares of the aquisition and analysis for SANS instrument must be developed and supplied in order to function properly. PS/PI block copolymer research in NIST will provide the general understanding of SANS instrument and instrument-related valuable informations such as standard sample for SANS and know-hows of the instrument building. The following are the results of this research. a. Construction of sample cell block. b. Software to control the temperature and auto-sample changer. c. Acquisition of the SANS data analysis routine and its modification for HANARO SANS. d. PS/PI block copolymer research in NIST. e. Calibration data of NIST and HANARO SANS for comparison. 39 figs., 2 tabs. (Author)

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Mohd, Shukri; Kassim, Razali; Mahmood, Zal Uyun [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Radiman, Shahidan

1998-10-01

The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

Incidência e fatores de risco da retinopatia diabética em pacientes do Hospital Universitário Onofre Lopes, Natal-RN Diabetic retinopathy incidence and risk factors in patients of the Onofre Lopes University Hospital, Natal-RN

Directory of Open Access Journals (Sweden)

Carlos Alexandre de Amorim Garcia

2003-06-01

Full Text Available OBJETIVO: Estudar a incidência e fatores de risco (tempo de doença e presença de hipertensão arterial sistêmica para retinopatia diabética em 1002 pacientes encaminhados pelo Programa de Diabetes do Hospital Universitário Onofre Lopes no período de 1992 - 1995. MÉTODOS: Estudo retrospectivo de pacientes com diagnóstico de diabetes mellitus encaminhados ao Setor de Retina do Departamento de Oftalmologia pelo Programa de Diabetes do Hospital Universitário e submetido, sob a supervisão do autor, a exame oftalmológico, incluindo medida da acuidade visual corrigida (tabela de Snellen, biomicroscopia do segmento anterior e posterior, tonometria de aplanação e oftalmoscopia binocular indireta sob midríase (tropicamida 1% + fenilefrina 10%. Foi realizada análise dos prontuários referente ao tempo de doenças e diagnostico clínico de hipertensão arterial sistêmica. RESULTADOS: Dos 1002 diabéticos examinados (em 24 deles a fundoscopia foi inviável, 978 foram separados em 4 grupos: sem retinopatia diabética (SRD, 675 casos (69,01%; com retinopatia diabética não proliferativa (RDNP, 207 casos (21,16%; com retinopatia diabética proliferativa (RDP, 70 casos (7,15%; e pacientes já fotocoagulados (JFC, 26 casos (2,65%. Do total, 291 eram do sexo masculino (29% e 711 do sexo feminino (71%. Os 4 grupos foram ainda avaliados quanto ao sexo, a faixa etária, a acuidade visual, tempo de doença, presença de catarata e hipertensão arterial sistêmica e comparados entre si. Com relação ao tipo de diabetes, 95 eram do tipo I (9,4%, 870 pacientes eram do tipo II (86,8%, e em 37 casos (3,7% o tipo de diabetes não foi determinado. CONCLUSÕES: Comprovou-se que os pacientes com maior tempo de doença tinham maior probabilidade de desenvolver retinopatia diabética, e que a hipertensão arterial sistêmica não constituiu fator de risco em relação à diminuição da acuidade visual nos pacientes hipertensos.PURPOSE: To study the incidence

Demonstration of reliability centered maintenance

International Nuclear Information System (INIS)

Schwan, C.A.; Morgan, T.A.

1991-04-01

Reliability centered maintenance (RCM) is an approach to preventive maintenance planning and evaluation that has been used successfully by other industries, most notably the airlines and military. Now EPRI is demonstrating RCM in the commercial nuclear power industry. Just completed are large-scale, two-year demonstrations at Rochester Gas ampersand Electric (Ginna Nuclear Power Station) and Southern California Edison (San Onofre Nuclear Generating Station). Both demonstrations were begun in the spring of 1988. At each plant, RCM was performed on 12 to 21 major systems. Both demonstrations determined that RCM is an appropriate means to optimize a PM program and improve nuclear plant preventive maintenance on a large scale. Such favorable results had been suggested by three earlier EPRI pilot studies at Florida Power ampersand Light, Duke Power, and Southern California Edison. EPRI selected the Ginna and San Onofre sites because, together, they represent a broad range of utility and plant size, plant organization, plant age, and histories of availability and reliability. Significant steps in each demonstration included: selecting and prioritizing plant systems for RCM evaluation; performing the RCM evaluation steps on selected systems; evaluating the RCM recommendations by a multi-disciplinary task force; implementing the RCM recommendations; establishing a system to track and verify the RCM benefits; and establishing procedures to update the RCM bases and recommendations with time (a living program). 7 refs., 1 tab

Recent development on Malaysian Small Angle Neutron Scattering (MySANS) facility upgrading and related research activities

International Nuclear Information System (INIS)

Abdul Aziz Bin Mohamed; Rafayudi Jamro; Razali Kassim; Muhammad Rawi Mat Zin; Azali Bin Muhammad; Muhd Noor Yunus; Dahlan Hj Mohd; Faridah Md Idris

2006-01-01

This paper describes the recent progress of the MySANS - 'mini' SANS facility and its present applications in materials science and technology research and education. Both management and technical strategies are generally explained. The formation of Reactor Interest Group (RIG) has lead to several experimental projects which collaborative work between MINT and local universities/research institutes. In addition a future work plan is also noted. (author)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

75 FR 12580 - Southern California Edison Company, San Onofre Nuclear Generating Station, Units 2 and 3...

Science.gov (United States)

2010-03-16

... exposures to plant workers and members of the public. Therefore, no changes or different types of.... There are no impacts to historical and cultural resources. There would be no impact to socioeconomic...

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

International Nuclear Information System (INIS)

Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

2013-01-01

The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

Nuclear power plant transient diagnostics using artificial neural networks that allow ''don't-know'' classifications

International Nuclear Information System (INIS)

Bartal, Y.; Lin, J.; Uhrig, R.E.

1995-01-01

A nuclear power plant's (NPP's) status is usually monitored by a human operator. Any classifier system used to enhance the operator's capability to diagnose a safety-critical system like an NPP should classify a novel transient as ''don't-know'' if it is not contained within its accumulated knowledge base. In particular, the classifier needs some kind of proximity measure between the new data and its training set. Artificial neural networks have been proposed as NPP classifiers, the most popular ones being the multilayered perceptron (MLP) type. However, MLPs do not have a proximity measure, while learning vector quantization, probabilistic neural networks (PNNs), and some others do. This proximity measure may also serve as an explanation to the classifier's decision in the way that case-based-reasoning expert systems do. The capability of a PNN network as a classifier is demonstrated using simulator data for the three-loop 436-MW(electric) Westinghouse San Onofre unit 1 pressurized water reactor. A transient's classification history is used in an ''evidence accumulation'' technique to enhance a classifier's accuracy as well as its consistency

Evaluation of computer-based NDE techniques and regional support of inspection activities

International Nuclear Information System (INIS)

Taylor, T.T.; Kurtz, R.J.; Heasler, P.G.; Doctor, S.R.

1991-01-01

This paper describes the technical progress during fiscal year 1990 for the program entitled 'Evaluation of Computer-Based nondestructive evaluation (NDE) Techniques and Regional Support of Inspection Activities.' Highlights of the technical progress include: development of a seminar to provide basic knowledge required to review and evaluate computer-based systems; review of a typical computer-based field procedure to determine compliance with applicable codes, ambiguities in procedure guidance, and overall effectiveness and utility; design and fabrication of a series of three test blocks for NRC staff use for training or audit of UT systems; technical assistance in reviewing (1) San Onofre ten year reactor pressure vessel inservice inspection activities and (2) the capability of a proposed phased array inspection of the feedwater nozzle at Oyster Creek; completion of design calculations to determine the feasibility and significance of various sizes of mockup assemblies that could be used to evaluate the effectiveness of eddy current examinations performed on steam generators; and discussion of initial mockup design features and methods for fabricating flaws in steam generator tubes

Nuclear Regulatory Commission issuances, April 1995. Volume 41, Number 4

International Nuclear Information System (INIS)

1995-04-01

This book contains issuances of the Nuclear Regulatory Commission and of the Atomic Safety and Licensing Boards, and an issuance of the Director's decision. The issuances concern a petition filed by Dr. James E Bauer seeking interlocutory Commission review of the Atomic Safety and Licensing Board's order imposing several restrictions on Dr. Bauer; a denial of an Interveners' Petition for Review addressing the application of Babcock and Wilcox for a renewal of its Special Nuclear Materials License; granting a motion for a protective order, by Sequoyah Fuel Corporation and General Atomics, limiting the use of the protected information to those individuals participating in the litigation and for the purposes of the litigation only; granting a Petitioner's petition for leave to intervene and request for a hearing concerning Georgia Institute of Technology (Georgia Tech Research Reactor) renewal of a facility license; and a denial of a petition filed by Mr. Ted Dougherty requesting a shutdown of the San Onofre Nuclear Generating Station based on concerns regarding the vulnerability of the plant to earthquakes and defensibility of the plant to a terrorist threat

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Massaro, Lawrence M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-09-30

A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power reactor sites was conducted. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

The Effect of Bangpungtongsung-san Extracts on Adipocyte Metabolism

Directory of Open Access Journals (Sweden)

Sang Min, Lee

2008-03-01

Full Text Available Objective : The purpose of this study is to investigate the effects of Bangpungtongsung-san extracts on the preadipocytes proliferation, of 3T3-L1 cell line. lipolysis of adipocytes in rat's epididymis and localized fat accumulation of porcine by extraction methods(alcohol and water. Methods : Diminish 3T3-L1 proliferation and lipogenesis do primary role to reduce obesity. So, 3T3-L1 preadipocyte and adipocytes were performed on cell cultures, and using Sprague-Dawley rats for the lipogenesis, and treated with 0.01-1 ㎎/㎖ Bangpungtongsung-san Extracts depend on concentrations. Porcine skin including fat tissue after treated Bangpungtongsung-san Extracts by means of the dosage dependent variation are investigated the histologic changes after injection of these extracts. Results : Following results were obtained from the 3T3-L1 preadipocyte proliferation and lipolysis of adipocyte in rats and histologic investigation of fat tissue. 1. Bangpungtongsung-san extracts were showed the effect of decreased preadipocyte proliferation on the high dosage(1.0㎎/㎖. 2. Bangpungtongsung-san extracts were showed the effect of decreased the activity of glycerol-3-phosphate dehydrogenase(GPDH on the high dosage(1.0㎎/㎖ and Specially, alcohol extract of Bangpungtongsung -san was clear as time goes by high concentration. 3. Bangpungtongsung-san extracts were showed tries to compare the effect of lipolysis, alcohol extract of Bangpungtongsung-san on the high dosage(1.0㎎/㎖ was observed the effect is higher than water extract. 4. Investigated the histological changes in porcine fat tissue after treated Bangpungtongsung-san extracts, we knew that water extract of Bangpungtongsung-san was showed the effect of lipolysis on the high dosage(10.0㎎/㎖ and alcohol extract of Bangpungtongsung-san was showed significant activity to the lysis of cell membranes in all concentration. Conclusion : These results suggest that Bangpungtongsung-san extracts efficiently

A study of copper precipitation in the thermally aged FeCu alloy using SANS

Energy Technology Data Exchange (ETDEWEB)

Park, D. G.; Kim, J. H.; Kwon, S. C.; Kim, W. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Lee, M. N.; Koo, Y. M. [Pohang University of Science and Technology, Pohang (Korea, Republic of)

2005-07-01

The continued operation or lifetime extension of a number of nuclear power plant around the world requires an understanding of the damage imparted to the reactor pressure vessel (RPV) steel by radiation. Irradiation embrittlement of nuclear reactor pressure vessel steels results from a high number of nanometer sized Cu rich precipitates (CRPs) and sub-nanometer defect-solute clusters. The copper precipitation leads to a distortion of the crystal lattice surrounding the copper precipitates and yields an internal micro-stress. In order to study the effect of copper precipitation on the steel embrittlement under neutron irradiation, the characteristics of nano size defects were investigated using small angle neutron scattering (SANS) in the thermal aged FeCu model alloys. The results on the precipitation composition, number density, size distribution and matrix composition obtained using a high resolution TEM and SANS are compared and contrasted.

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

A new small-angle neutron scattering spectrometer at China Mianyang research reactor

Energy Technology Data Exchange (ETDEWEB)

Peng, Mei, E-mail: pm740509@163.com; Sun, Liangwei; Chen, Liang; Sun, Guangai; Chen, Bo; Xie, Chaomei; Xia, Qingzhong; Yan, Guanyun; Tian, Qiang; Huang, Chaoqiang; Pang, Beibei; Zhang, Ying; Wang, Yun; Liu, Yaoguang; Kang, Wu; Gong, Jian

2016-02-21

A new pinhole small-angle neutron scattering (SANS) spectrometer, installed at the cold neutron source of the 20 MW China Mianyang Research Reactor (CMRR) in the Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, has been put into use since 2014. The spectrometer is equipped with a multi-blade mechanical velocity selector, a multi-beam collimation system, and a two-dimensional He-3 position sensitive neutron detector. The q-range of the spectrometer covers from 0.01 nm{sup −1} to 5.0 nm{sup −1}. In this paper, the design and characteristics of the SANS spectrometer are described. The q-resolution calculations, together with calibration measurements of silver behenate and a dispersion of nearly monodisperse poly-methyl-methacrylate nanoparticles indicate that our SANS spectrometer has a good performance and is now in routine service. - Highlights: • A new SANS spectrometer has been put into use since 2014 in China. • One MBR selector possesses a higher resolution compared with traditional selector is used. • The spectrometer has a good performance and is now in routinely service.

Removal in a lump of JRR-3 nuclear reactor

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

1989-01-01

The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

G2 and G3 reactors design; Description des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Combination of SANS and 3D stochastic reconstruction techniques for the study of nanostructured materials

CERN Document Server

Kikkinides, E S; Steriotis, T A; Kanellopoulos, N K; Mitropoulos, A C; Treimer, W

2002-01-01

Ceramic nanostructured materials have recently received scientific and industrial interest due to their unique properties. A series of such nanoporous structures were characterised by SANS techniques. The resulting scattering curves were analysed to obtain basic structural information regarding the pore size distribution and autocorrelation function of each material. Furthermore, stochastic reconstruction models were employed to generate 3D images with the same basic structural characteristics obtained from SANS. Finally, simulation results of permeation on the reconstructed images provide very good agreement with experimental data. (orig.)

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

Electricity-market price and nuclear power plant shutdown: Evidence from California

International Nuclear Information System (INIS)

Woo, C.K.; Ho, T.; Zarnikau, J.; Olson, A.; Jones, R.; Chait, M.; Horowitz, I.; Wang, J.

2014-01-01

Japan's Fukushima nuclear disaster, triggered by the March 11, 2011 earthquake, has led to calls for shutting down existing nuclear plants. To maintain resource adequacy for a grid's reliable operation, one option is to expand conventional generation, whose marginal unit is typically fueled by natural-gas. Two timely and relevant questions thus arise for a deregulated wholesale electricity market: (1) what is the likely price increase due to a nuclear plant shutdown? and (2) what can be done to mitigate the price increase? To answer these questions, we perform a regression analysis of a large sample of hourly real-time electricity-market price data from the California Independent System Operator (CAISO) for the 33-month sample period of April 2010–December 2012. Our analysis indicates that the 2013 shutdown of the state's San Onofre plant raised the CAISO real-time hourly market prices by $6/MWH to $9/MWH, and that the price increases could have been offset by a combination of demand reduction, increasing solar generation, and increasing wind generation. - Highlights: • Japan's disaster led to calls for shutting down existing nuclear plants. • We perform a regression analysis of California's real-time electricity-market prices. • We estimate that the San Onofre plant shutdown has raised the market prices by $6/MWH to $9/MWH. • The price increases could be offset by demand reduction and renewable generation increase

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Genetic analysis of the spindle checkpoint genes san-1, mdf-2, bub-3 and the CENP-F homologues hcp-1 and hcp-2 in Caenorhabditis elegans

Directory of Open Access Journals (Sweden)

Moore Landon L

2008-02-01

Full Text Available Abstract Background The spindle checkpoint delays the onset of anaphase until all sister chromatids are aligned properly at the metaphase plate. To investigate the role san-1, the MAD3 homologue, has in Caenorhabditis elegans embryos we used RNA interference (RNAi to identify genes synthetic lethal with the viable san-1(ok1580 deletion mutant. Results The san-1(ok1580 animal has low penetrating phenotypes including an increased incidence of males, larvae arrest, slow growth, protruding vulva, and defects in vulva morphogenesis. We found that the viability of san-1(ok1580 embryos is significantly reduced when HCP-1 (CENP-F homologue, MDF-1 (MAD-1 homologue, MDF-2 (MAD-2 homologue or BUB-3 (predicted BUB-3 homologue are reduced by RNAi. Interestingly, the viability of san-1(ok1580 embryos is not significantly reduced when the paralog of HCP-1, HCP-2, is reduced. The phenotype of san-1(ok1580;hcp-1(RNAi embryos includes embryonic and larval lethality, abnormal organ development, and an increase in abnormal chromosome segregation (aberrant mitotic nuclei, anaphase bridging. Several of the san-1(ok1580;hcp-1(RNAi animals displayed abnormal kinetochore (detected by MPM-2 and microtubule structure. The survival of mdf-2(RNAi;hcp-1(RNAi embryos but not bub-3(RNAi;hcp-1(RNAi embryos was also compromised. Finally, we found that san-1(ok1580 and bub-3(RNAi, but not hcp-1(RNAi embryos, were sensitive to anoxia, suggesting that like SAN-1, BUB-3 has a functional role as a spindle checkpoint protein. Conclusion Together, these data suggest that in the C. elegans embryo, HCP-1 interacts with a subset of the spindle checkpoint pathway. Furthermore, the fact that san-1(ok1580;hcp-1(RNAi animals had a severe viability defect whereas in the san-1(ok1580;hcp-2(RNAi and san-1(ok1580;hcp-2(ok1757 animals the viability defect was not as severe suggesting that hcp-1 and hcp-2 are not completely redundant.

75 FR 71152 - Southern California Edison; San Onofre Nuclear Generating Station, Unit 2 and Unit 3; Exemption

Science.gov (United States)

2010-11-22

... requirements of this section through its Commission-approved Physical Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Cyber Security Plan referred to collectively hereafter as...

Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960); Etude d'un element combustible en oxyde d'uranium gaine d'aluminium, type ''cristal de neige'' pour la pile EL 3 (1960)

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M; Guibert, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [French] Ce rapport resume les principales etudes qui ont ete faites sur l'element combustible 'Cristal de Neige' (a oxyde d'uranium, gaine d'aluminium) destine a remnlacer l'element actuel du reacteur experimental EL3, afin d'en augmenter la reactivite sans modifier le rapport flux neutronique-puissance thermique. (auteur)

75 FR 38412 - Safety Zone; San Diego POPS Fireworks, San Diego, CA

Science.gov (United States)

2010-07-02

...-AA00 Safety Zone; San Diego POPS Fireworks, San Diego, CA AGENCY: Coast Guard, DHS. ACTION: Temporary... waters of San Diego Bay in support of the San Diego POPS Fireworks. This safety zone is necessary to... San Diego POPS Fireworks, which will include fireworks presentations conducted from a barge in San...

San Marco C-2 (San Marco-4) Post Launch Report No. 1

Science.gov (United States)

1974-01-01

The San Marco C-2 spacecraft, now designated San Marco-4, was successfully launched by a Scout vehicle from the San Marco Platform on 18 February 1974 at 6:05 a.m. EDT. The launch occurred 2 hours 50 minutes into the 3-hour window due co low cloud cover at the launch site. All spacecraft subsystems have been checked and are functioning normally. The protective caps for the two U.S. experiments were ejected and the Omegatron experiment activated on 19 February. The neutral mass spectrometer was activated as scheduled on 22 February after sufficient time to allow for spacecraft outgassing and to avoid the possibility of corona occurring. Both instruments are performing properly and worthwhile scientific data is being acquired.

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

D-3He fueled FRC reactor 'ARTEMIS-L'

International Nuclear Information System (INIS)

Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

1992-09-01

A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

Science.gov (United States)

2010-09-17

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0178; Docket No. 50-228; License No. R-98] In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order Extending the... possession, use, and operation of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon...

75 FR 39985 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

Science.gov (United States)

2010-07-13

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0178; Docket No. 50-228; License No. R-98] In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order Approving Indirect... of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon, California, under the...

Description of the RA-3 research reactor as a model facility

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

Nuclear Regulatory Commission issuances, April 1995. Volume 41, Number 4

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-04-01

This book contains issuances of the Nuclear Regulatory Commission and of the Atomic Safety and Licensing Boards, and an issuance of the Director`s decision. The issuances concern a petition filed by Dr. James E Bauer seeking interlocutory Commission review of the Atomic Safety and Licensing Board`s order imposing several restrictions on Dr. Bauer; a denial of an Interveners` Petition for Review addressing the application of Babcock and Wilcox for a renewal of its Special Nuclear Materials License; granting a motion for a protective order, by Sequoyah Fuel Corporation and General Atomics, limiting the use of the protected information to those individuals participating in the litigation and for the purposes of the litigation only; granting a Petitioner`s petition for leave to intervene and request for a hearing concerning Georgia Institute of Technology (Georgia Tech Research Reactor) renewal of a facility license; and a denial of a petition filed by Mr. Ted Dougherty requesting a shutdown of the San Onofre Nuclear Generating Station based on concerns regarding the vulnerability of the plant to earthquakes and defensibility of the plant to a terrorist threat.

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

78 FR 19103 - Safety Zone; Spanish Navy School Ship San Sebastian El Cano Escort; Bahia de San Juan; San Juan, PR

Science.gov (United States)

2013-03-29

...-AA00 Safety Zone; Spanish Navy School Ship San Sebastian El Cano Escort; Bahia de San Juan; San Juan... temporary moving safety zone on the waters of Bahia de San Juan during the transit of the Spanish Navy... Channel entrance, and to protect the high ranking officials on board the Spanish Navy School Ship San...

Digital upgrade of radiation-monitoring-system subcomponents

International Nuclear Information System (INIS)

Bohrisch, R.L

1993-01-01

This paper describes the experience of Southern California Edison (SCE) in upgrading an obsolete, analog, printed circuit board contain in most of the process and effluent radiation detectors at the San Onofre Nuclear Generating Station. The printed circuit board, which functions to produce a linear voltage and current that is proportional to the log of the radiation level, was reengineered by SCE with microprocessor-based digital technology and subjected to qualification testing, including seismic and environmental, for use in class I safety-related applications. The results, benefits, and disadvantages to this approach are discussed in this paper

Design requirements, operation and maintenance of gas-cooled reactors

International Nuclear Information System (INIS)

1989-06-01

At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

2016-09-01

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

76 FR 45693 - Safety Zone; San Diego POPS Fireworks, San Diego, CA

Science.gov (United States)

2011-08-01

...-AA00 Safety Zone; San Diego POPS Fireworks, San Diego, CA AGENCY: Coast Guard, DHS. ACTION: Temporary... San Diego Bay in support of the San Diego POPS Fireworks. This safety zone is necessary to provide for... of the waterway during scheduled fireworks events. Persons and vessels will be prohibited from...

Investigations on neutron irradiated 3D carbon fibre reinforced carbon composite material

Science.gov (United States)

Venugopalan, Ramani; Alur, V. D.; Patra, A. K.; Acharya, R.; Srivastava, D.

2018-04-01

As against conventional graphite materials carbon-carbon (C/C) composite materials are now being contemplated as the promising candidate materials for the high temperature and fusion reactor owing to their high thermal conductivity and high thermal resistance, better mechanical/thermal properties and irradiation stability. The current need is for focused research on novel carbon materials for future new generation nuclear reactors. The advantage of carbon-carbon composite is that the microstructure and the properties can be tailor made. The present study encompasses the irradiation of 3D carbon composite prepared by reinforcement using PAN carbon fibers for nuclear application. The carbon fiber reinforced composite was subjected to neutron irradiation in the research reactor DHRUVA. The irradiated samples were characterized by Differential Scanning Calorimetry (DSC), small angle neutron scattering (SANS), XRD and Raman spectroscopy. The DSC scans were taken in argon atmosphere under a linear heating program. The scanning was carried out at temperature range from 30 °C to 700 °C at different heating rates in argon atmosphere along with reference as unirradiated carbon composite. The Wigner energy spectrum of irradiated composite showed two peaks corresponding to 200 °C and 600 °C. The stored energy data for the samples were in the range 110-170 J/g for temperature ranging from 30 °C to 700 °C. The Wigner energy spectrum of irradiated carbon composite did not indicate spontaneous temperature rise during thermal annealing. Small angle neutron scattering (SANS) experiments have been carried out to investigate neutron irradiation induced changes in porosity of the composite samples. SANS data were recorded in the scattering wave vector range of 0.17 nm-1 to 3.5 nm-1. Comparison of SANS profiles of irradiated and unirradiated samples indicates significant change in pore morphology. Pore size distributions of the samples follow power law size distribution with

33 CFR 165.754 - Safety Zone: San Juan Harbor, San Juan, PR.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety Zone: San Juan Harbor, San Juan, PR. 165.754 Section 165.754 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND... Zone: San Juan Harbor, San Juan, PR. (a) Regulated area. A moving safety zone is established in the...

ASTER Flyby of San Francisco

Science.gov (United States)

2002-01-01

The Advanced Spaceborne Thermal Emission and Reflection radiometer, ASTER, is an international project: the instrument was supplied by Japan's Ministry of International Trade and Industry. A joint US/Japan science team developed algorithms for science data products, and is validating instrument performance. With its 14 spectral bands, extremely high spatial resolution, and 15 meter along-track stereo capability, ASTER is the zoom lens of the Terra satellite. The primary mission goals are to characterize the Earth's surface; and to monitor dynamic events and processes that influence habitability at human scales. ASTER's monitoring and mapping capabilities are illustrated by this series of images of the San Francisco area. The visible and near infrared image reveals suspended sediment in the bays, vegetation health, and details of the urban environment. Flying over San Francisco (3.2MB) (high-res (18.3MB)), we see the downtown, and shadows of the large buildings. Past the Golden Gate Bridge and Alcatraz Island, we cross San Pablo Bay and enter Suisun Bay. Turning south, we fly over the Berkeley and Oakland Hills. Large salt evaporation ponds come into view at the south end of San Francisco Bay. We turn northward, and approach San Francisco Airport. Rather than landing and ending our flight, we see this is as only the beginning of a 6 year mission to better understand the habitability of the world on which we live. For more information: ASTER images through Visible Earth ASTER Web Site Image courtesy of MITI, ERSDAC, JAROS, and the U.S./Japan ASTER Science Team

78 FR 27260 - Southern California Edison, San Onofre Nuclear Generating Station, Units 2 and 3 Request for Action

Science.gov (United States)

2013-05-09

... (NRC) is giving notice that by petition dated June 18, 2012, Friends of the Earth (FOE, the petitioner... 12, 2013 (ADAMS Accession No. ML13116A265), FOE requested that Mitsubishi Heavy Industries' Report... this petition within a reasonable time. Further, FOE submitted on April 4, 2013, a cover letter and...

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

Gas-cooled reactor coolant circulator and blower technology

International Nuclear Information System (INIS)

1988-08-01

In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

2014-01-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

Science.gov (United States)

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

1993-11-01

The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

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

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

Coal exploration in the Alto San Jorge area, Cordoba Department. Exploracion de carbones en el Ato San Jorge, Departamento de Cordoba

Energy Technology Data Exchange (ETDEWEB)

Ospina, L H; Oquendo, G G [Geominas Ltda, Medellin (Colombia)

1989-01-01

A Mining Feasibility Study in the Area of Alto San Jorge, Department of Cordoba, Colombia, was commissioned by CARBOCOL S.A. to the Consortium Geominas-NACI. An area of 800 Ka2 was explored to define surface mining possibilities within two subareas referred to as Alto San Jorge and San Pedro Ure. Rocks of Cretaceous, Tertiary and Quaternary age crop out in the zone. In the subarea Alto San Jorge the principal structure is a syncline with a south-north direction. The San Pedro Ure subarea is formed by undulations with flanks of low dip, the most important being the San Antonio Syncline because it contains the mining block. The geological study of the surface demonstrated the existence of coal in the Oligocene Cienaga de Oro Formation and the Niocene Cerrito Formation, with potential resources of 6.3 billion tons. The subsequent exploration of the subsoil, with 20.618 m of drilling, permitted determination of demonstrated reserves in the order of 2.9 billion tons within two areas. In the sector selected for the mine plan, in the area of San Pedro-Puerto Libertador, 7.791 m of drilling was accomplished to define a demonstrated reserve of 515 million tons of coal down to a depth of 200. The combustible type coal has 5.000 cal/g. Complete mining schedules were developed at the prefeasibility level for two surface mines with productions of 1.5 MMTY and 4 MMTY. 9 figs., 3 tabs., 28 refs.

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

San Onofre Nuclear Generating Station, Unit 1. Annual operating report for 1976

International Nuclear Information System (INIS)

1977-01-01

Gross electrical energy generated was 2,610,000 MWH with the generator on line 6,162.9 hrs. Information is presented concerning operations, power generation, shutdowns, corrective maintenance, chemistry and radiochemistry, occupational radiation exposure, release of radioactive materials, reportable occurrences, steam generator tube inspections, primary coolant chemistry, containment penetration leak tests, and radiological environmental monitoring

Design of a TOF-SANS instrument for the proposed long wavelength target station at the spallation neutron source

International Nuclear Information System (INIS)

Thiyagarajan, P.; Littrell, K.; Seeger, P.A.

2001-01-01

We have designed a versatile high-throughput SANS instrument [Broad Range Intense Multipurpose SANS (BRIMS)] for the proposed Long Wavelength Target Station at the SNS by using acceptance diagrams and the Los Alamos NISP Monte Carlo simulation package. This instrument has been fully optimized to take advantage of the 10 Hz source frequency (broad wavelength bandwidth) and the cold neutron spectrum from a tall coupled solid methane moderator (12 cm x 20 cm). BRIMS has been designed to produce data in a Q range spanning from 0.0025 to 0.7 A -1 in a single measurement by simultaneously using neutrons with wavelengths ranging from 1 to 14.5 A in a time of flight mode. A supermirror guide and bender assembly is employed to separate and redirect the useful portion of the neutron spectrum with λ>1 A, by 2.3deg away from the direct beam containing high energy neutrons and γ rays. The effects of various collimation choices on count rate, resolution and Q min have been characterized using spherical particle and delta function scatterers. The overall performance of BRIMS has been compared with that of the best existing reactor-based SANS instrument D22 at ILL. (author)

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

Characteristics of D-3He fueled frc reactor: ARTEMIS-L

International Nuclear Information System (INIS)

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

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Minería, conflicto y mediadores locales: Minera San Xavier en Cerro de San Pedro, México Mineira, conflito e mediadores locais: Minera San Xavier em Cerro de San Pedro Mining, conflict and local brokers: Minera San Xavier in Cerro de San Pedro

Directory of Open Access Journals (Sweden)

Hernán Horacio Schiaffini

2011-12-01

Full Text Available Este trabajo indaga en las instancias de mediación que intervienen en la articulación de procesos económicos de gran escala y su puesta en práctica local. Basándonos en el conflicto que se produjo en el Municipio de Cerro de San Pedro (San Luis Potosí, México entre la empresa Minera San Xavier y el Frente Amplio Opositor (FAO a la misma, aplicamos el método etnográfico con el objetivo de describir las estructuras locales de mediación política y analizar sus prácticas y racionalidad. Intentamos demostrar así la importancia de los factores políticos locales en las vinculaciones entre estado, empresa y población.Este trabalho indaga nas instâncias de mediação que intervêm em processos econômicos de grande escala e sua posta em prática local. Baseando-nos no conflito no Cerro de San Pedro (San Luis Potosí, México entre a empresa Minera San Xavier e a Frente Amplio Opositor (FAO aplicamos o método etnográfico pra descrever as estruturas de mediação política locais e analisar suas práticas e racionalidade. Tenta-se demonstrar assim a importância dos fatores políticos locais nas vinculações entre estado, empresa e população.This paper investigates in instances of mediation involved in large-scale economic processes and local implementation. Analyzing the conflict in Cerro de San Pedro (San Luis Potosí, México among San Xavier mining company and the Frente Amplio Opositor (FAO, it applies an ethnographic approach to describe the local structures of political mediation and its practices and rationality. The work shows the relevance of local factors in the relationships between State, company and people.

Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960); Etude d'un element combustible en oxyde d'uranium gaine d'aluminium, type ''cristal de neige'' pour la pile EL 3 (1960)

Energy Technology Data Exchange (ETDEWEB)

Gauthron, M.; Guibert, B. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [French] Ce rapport resume les principales etudes qui ont ete faites sur l'element combustible 'Cristal de Neige' (a oxyde d'uranium, gaine d'aluminium) destine a remnlacer l'element actuel du reacteur experimental EL3, afin d'en augmenter la reactivite sans modifier le rapport flux neutronique-puissance thermique. (auteur)

Distribution and demography of San Francisco gartersnakes (Thamnophis sirtalis tetrataenia) at Mindego Ranch, Russian Ridge Open Space Preserve, San Mateo County, California

Science.gov (United States)

Kim, Richard; Halstead, Brian J.; Wylie, Glenn D.; Casazza, Michael L.

2018-04-26

San Francisco gartersnakes (Thamnophis sirtalis tetrataenia) are a subspecies of common gartersnakes endemic to the San Francisco Peninsula of northern California. Because of habitat loss and collection for the pet trade, San Francisco gartersnakes were listed as endangered under the precursor to the Federal Endangered Species Act. A population of San Francisco gartersnakes resides at Mindego Ranch, San Mateo County, which is part of the Russian Ridge Open Space Preserve owned and managed by the Midpeninsula Regional Open Space District (MROSD). Because the site contained non-native fishes and American bullfrogs (Lithobates catesbeianus), MROSD implemented management to eliminate or reduce the abundance of these non-native species in 2014. We monitored the population using capture-mark-recapture techniques to document changes in the population during and following management actions. Although drought confounded some aspects of inference about the effects of management, prey and San Francisco gartersnake populations generally increased following draining of Aquatic Feature 3. Continued management of the site to keep invasive aquatic predators from recolonizing or increasing in abundance, as well as vegetation management that promotes heterogeneous grassland/shrubland near wetlands, likely would benefit this population of San Francisco gartersnakes.

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

76 FR 1386 - Safety Zone; Centennial of Naval Aviation Kickoff, San Diego Bay, San Diego, CA

Science.gov (United States)

2011-01-10

...-AA00 Safety Zone; Centennial of Naval Aviation Kickoff, San Diego Bay, San Diego, CA AGENCY: Coast... zone on the navigable waters of San Diego Bay in San Diego, CA in support of the Centennial of Naval... February 12, 2010, the Centennial of Naval Aviation Kickoff will take place in San Diego Bay. In support of...

Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1996-07-01

The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

San Marino.

Science.gov (United States)

1985-02-01

San Marino, an independent republic located in north central Italy, in 1983 had a population of 22,206 growing at an annual rate of .9%. The literacy rate is 97% and the infant mortality rate is 9.6/1000. The terrain is mountainous and the climate is moderate. According to local tradition, San Marino was founded by a Christian stonecutter in the 4th century A.D. as a refuge against religious persecution. Its recorded history began in the 9th century, and it has survived assaults on its independence by the papacy, the Malatesta lords of Rimini, Cesare Borgia, Napoleon, and Mussolini. An 1862 treaty with the newly formed Kingdom of Italy has been periodically renewed and amended. The present government is an alliance between the socialists and communists. San Marino has had its own statutes and governmental institutions since the 11th century. Legislative authority at present is vested in a 60-member unicameral parliament. Executive authority is exercised by the 11-member Congress of State, the members of which head the various administrative departments of the goverment. The posts are divided among the parties which form the coalition government. Judicial authority is partly exercised by Italian magistrates in civil and criminal cases. San Marino's policies are tied to Italy's and political organizations and labor unions active in Italy are also active in San Marino. Since World War II, there has been intense rivalry between 2 political coalitions, the Popular Alliance composed of the Christian Democratic Party and the Independent Social Democratic Party, and the Liberty Committee, coalition of the Communist Party and the Socialist Party. San Marino's gross domestic product was $137 million and its per capita income was $6290 in 1980. The principal economic activities are farming and livestock raising, along with some light manufacturing. Foreign transactions are dominated by tourism. The government derives most of its revenue from the sale of postage stamps to

San Francisco District Laboratory (SAN)

Data.gov (United States)

Federal Laboratory Consortium — Program CapabilitiesFood Analysis SAN-DO Laboratory has an expert in elemental analysis who frequently performs field inspections of materials. A recently acquired...

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Quaternary geology of Alameda County, and parts of Contra Costa, Santa Clara, San Mateo, San Francisco, Stanislaus, and San Joaquin counties, California: a digital database

Science.gov (United States)

Helley, E.J.; Graymer, R.W.

1997-01-01

Alameda County is located at the northern end of the Diablo Range of Central California. It is bounded on the north by the south flank of Mount Diablo, one of the highest peaks in the Bay Area, reaching an elevation of 1173 meters (3,849 ft). San Francisco Bay forms the western boundary, the San Joaquin Valley borders it on the east and an arbitrary line from the Bay into the Diablo Range forms the southern boundary. Alameda is one of the nine Bay Area counties tributary to San Francisco Bay. Most of the country is mountainous with steep rugged topography. Alameda County is covered by twenty-eight 7.5' topographic Quadrangles which are shown on the index map. The Quaternary deposits in Alameda County comprise three distinct depositional environments. One, forming a transgressive sequence of alluvial fan and fan-delta facies, is mapped in the western one-third of the county. The second, forming only alluvial fan facies, is mapped in the Livermore Valley and San Joaquin Valley in the eastern part of the county. The third, forming a combination of Eolian dune and estuarine facies, is restricted to the Alameda Island area in the northwestern corner of the county.

Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

76 FR 9709 - Water Quality Challenges in the San Francisco Bay/Sacramento-San Joaquin Delta Estuary

Science.gov (United States)

2011-02-22

... Water Quality Challenges in the San Francisco Bay/Sacramento-San Joaquin Delta Estuary AGENCY... the San Francisco Bay/ Sacramento-San Joaquin Delta Estuary (Bay Delta Estuary) in California. EPA is... programs to address recent significant declines in multiple aquatic species in the Bay Delta Estuary. EPA...

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Modeling pesticide loadings from the San Joaquin watershed into the Sacramento-San Joaquin Delta using SWAT

Science.gov (United States)

Chen, H.; Zhang, M.

2016-12-01

The Sacramento-San Joaquin Delta is an ecologically rich, hydrologically complex area that serves as the hub of California's water supply. However, pesticides have been routinely detected in the Delta waterways, with concentrations exceeding the benchmark for the protection of aquatic life. Pesticide loadings into the Delta are partially attributed to the San Joaquin watershed, a highly productive agricultural watershed located upstream. Therefore, this study aims to simulate pesticide loadings to the Delta by applying the Soil and Water Assessment Tool (SWAT) model to the San Joaquin watershed, under the support of the USDA-ARS Delta Area-Wide Pest Management Program. Pesticide use patterns in the San Joaquin watershed were characterized by combining the California Pesticide Use Reporting (PUR) database and GIS analysis. Sensitivity/uncertainty analyses and multi-site calibration were performed in the simulation of stream flow, sediment, and pesticide loads along the San Joaquin River. Model performance was evaluated using a combination of graphic and quantitative measures. Preliminary results indicated that stream flow was satisfactorily simulated along the San Joaquin River and the major eastern tributaries, whereas stream flow was less accurately simulated in the western tributaries, which are ephemeral small streams that peak during winter storm events and are mainly fed by irrigation return flow during the growing season. The most sensitive parameters to stream flow were CN2, SOL_AWC, HRU_SLP, SLSUBBSN, SLSOIL, GWQMN and GW_REVAP. Regionalization of parameters is important as the sensitivity of parameters vary significantly spatially. In terms of evaluation metric, NSE tended to overrate model performance when compared to PBIAS. Anticipated results will include (1) pesticide use pattern analysis, (2) calibration and validation of stream flow, sediment, and pesticide loads, and (3) characterization of spatial patterns and temporal trends of pesticide yield.

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

Online monitoring and diagnostic system on RA-6 nuclear reactor

International Nuclear Information System (INIS)

Garcia Peyrano, O. A.; Marticorena, M.; Koch, R. G.; Martinez, J. S; Berruti, G. E.; Nunez, W. M.; Gonzales, L. A.; Tarquini, L. D.; Sotelo, J. P

2009-01-01

This paper presents the Online Automatic Monitoring and Diagnostic System for mechanical components, installed on RA-6 Nuclear Reactor (San Carlos de Bariloche, Argentina). This system has been designed, installed and set-up by the Vibrations and Mechatronics Laboratory (Centro Atomico Bariloche, Comision Nacional de Energia Atomica) and Sitrack.com Argentina SA. This system provides an online mechanical diagnostic of the main reactor components, allowing incipient failures to be early detected and identified, avoiding unscheduled shut-downs and reducing maintenance times. The diagnostic is accomplished by an online analysis of the vibratory signature of the mechanical components, obtained by vibrations sensors on the main pump and the decay tank. The mechanical diagnostic and the main operational parameters are displayed on the reactor control room and published on the internet. [es

33 CFR 165.776 - Security Zone; Coast Guard Base San Juan, San Juan Harbor, Puerto Rico

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Security Zone; Coast Guard Base San Juan, San Juan Harbor, Puerto Rico 165.776 Section 165.776 Navigation and Navigable Waters COAST... Guard District § 165.776 Security Zone; Coast Guard Base San Juan, San Juan Harbor, Puerto Rico (a...

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Managing the systems approach to training using a flexible Hierarchical data base

International Nuclear Information System (INIS)

Housman, E.; Bush, E.R.

1993-01-01

Task analysis/curriculum design for a nuclear power station results in a massive amount of data, which must be sequenced and ordered to create an effective program design. This is an almost impossible task without the use of computerized data base. Beginning in 1989, San Onofre nuclear generating station (SONGS) undertook a task analysis/program design project to verify the structure and sequence (design) of all accredited training program. A flex hierarchical data-base management system was designed to store and manage the data collected during the project. For the Operations Training Programm alone ∼8000 tasks, 90,000 knowledges and abilities, and 10,000 learning objectives were entered into this data base

Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor; Essais de transmission de chaleur sur maquette pour l'etude du refroidissement de la pile EL 3

Energy Technology Data Exchange (ETDEWEB)

Raievski, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Butzbach, M; Domenjoud, M [Alsthom, 75 - Paris (France); Bousquet, M [Chantiers de l' Atlantique (France); Braudeau, M; Milliat, M [Electricite de France (EDF), 75 - Paris (France)

1958-07-01

For such high heat flux density as is released in the channels of EL3 reactor (2.10{sup 6} kcal/m{sup 2}h on the hottest point) cooling conditions have proved to be satisfactory, that is free from nucleate boiling. The arrangements provided for these tests and the technique used for measurements (of temperature particularly) are specified. Two fields have been investigated: in the former (forced convection without nucleate boiling) a good agreement is found with Colburn's formula. The influence of the ratio L/D is pointed out. The latter field is of forced convection with beginning of nucleate boiling; there the observed raise of the transfer coefficient has been shown occurring with some delay. (author)Fren. [French] A la valeur elevee prevue pour la densite de flux de chaleur (2.10{sup 6} kcal/m{sup 2}h au point le plus chaud) il est verifie que le refroidissement de la pile s'effectue normalement (sans ebullition de paroi). Les essais sont menes sur la maquette grandeur nature d'un canal d'EL3. Les dispositions relatives a la conduite des essais et a la technique des mesures (de temperature en particulier) sont precisees. Deux domaines sont etudies; pour t{sub p} < T{sub sat} (convection forcee sans ebullition de paroi) on constate un bon accord avec la formule de Colburn, avec toutefois l'influence du rapport L/D. Pour t{sub p} < T{sub sat} (debut d'ebullition) l'augmentation prevue du coefficient de transmission presente un certain retard. (auteur)

Enhanced Preliminary Assessment Report: Presidio of San Francisco Military Reservation, San Francisco, California

Science.gov (United States)

1989-11-01

CAD981415656 Filmore Steiner Bay San Francisco 24 PG&E Gas Plant SanFran 502-IG CAD981415714 Bay North Point Buchanan Laguna 25 PG&E Gas Plant SanFran 502-1H...76-ioV /5,JO /0.7 /,230 PSF Water PSF, Main U.N. Lagunda Honda Analvte Plant Clearwell Reservoir Plaza Reservoi- Chlordane inetab. ə.2 ə.2 (1.2 ə.2

Development of an updated fundamental basic wind speed map for SANS 10160-3

CSIR Research Space (South Africa)

Kruger, A

2017-12-01

Full Text Available winds is of car- dinal importance to the built environment, and should be updated as new information becomes available. A review of the historical development of climatic data for wind load design in South Africa is provided by Goliger et al (2017..., Goliger AM. Development of an updated fundamental basic wind speed map for SANS 10160-3. J. S. Afr. Inst. Civ. Eng. 2017:59(4), Art. #1739, 14 pages. http://dx.doi.org/10.17159/2309-8775/2017/v59n4a2 TECHNICAL PAPER Journal of the South african in...

Estimating natural recharge in San Gorgonio Pass watersheds, California, 1913–2012

Science.gov (United States)

Hevesi, Joseph A.; Christensen, Allen H.

2015-12-21

A daily precipitation-runoff model was developed to estimate spatially and temporally distributed recharge for groundwater basins in the San Gorgonio Pass area, southern California. The recharge estimates are needed to define transient boundary conditions for a groundwater-flow model being developed to evaluate the effects of pumping and climate on the long-term availability of groundwater. The area defined for estimating recharge is referred to as the San Gorgonio Pass watershed model (SGPWM) and includes three watersheds: San Timoteo Creek, Potrero Creek, and San Gorgonio River. The SGPWM was developed by using the U.S. Geological Survey INFILtration version 3.0 (INFILv3) model code used in previous studies of recharge in the southern California region, including the San Gorgonio Pass area. The SGPWM uses a 150-meter gridded discretization of the area of interest in order to account for spatial variability in climate and watershed characteristics. The high degree of spatial variability in climate and watershed characteristics in the San Gorgonio Pass area is caused, in part, by the high relief and rugged topography of the area.

A case for historic joint rupture of the San Andreas and San Jacinto faults.

Science.gov (United States)

Lozos, Julian C

2016-03-01

The San Andreas fault is considered to be the primary plate boundary fault in southern California and the most likely fault to produce a major earthquake. I use dynamic rupture modeling to show that the San Jacinto fault is capable of rupturing along with the San Andreas in a single earthquake, and interpret these results along with existing paleoseismic data and historic damage reports to suggest that this has likely occurred in the historic past. In particular, I find that paleoseismic data and historic observations for the ~M7.5 earthquake of 8 December 1812 are best explained by a rupture that begins on the San Jacinto fault and propagates onto the San Andreas fault. This precedent carries the implications that similar joint ruptures are possible in the future and that the San Jacinto fault plays a more significant role in seismic hazard in southern California than previously considered. My work also shows how physics-based modeling can be used for interpreting paleoseismic data sets and understanding prehistoric fault behavior.

A case for historic joint rupture of the San Andreas and San Jacinto faults

Science.gov (United States)

Lozos, Julian C.

2016-01-01

The San Andreas fault is considered to be the primary plate boundary fault in southern California and the most likely fault to produce a major earthquake. I use dynamic rupture modeling to show that the San Jacinto fault is capable of rupturing along with the San Andreas in a single earthquake, and interpret these results along with existing paleoseismic data and historic damage reports to suggest that this has likely occurred in the historic past. In particular, I find that paleoseismic data and historic observations for the ~M7.5 earthquake of 8 December 1812 are best explained by a rupture that begins on the San Jacinto fault and propagates onto the San Andreas fault. This precedent carries the implications that similar joint ruptures are possible in the future and that the San Jacinto fault plays a more significant role in seismic hazard in southern California than previously considered. My work also shows how physics-based modeling can be used for interpreting paleoseismic data sets and understanding prehistoric fault behavior. PMID:27034977

Project management with regulatory constraints: Return to service of San Onofre Unit 1

International Nuclear Information System (INIS)

Hosmer, J.

1986-01-01

Station acceptance of the RTS work was completed in early November, 1984. The unit completed a successful integrated leak rate test, hot function test, and 200 hour warranty run in December, 1984 six weeks ahead of the CPUC mandate. Capital expenditures, although greater than the initial forecast, were two million dollars less than the 37.5 million dollar CPUC cap. The unit has run essentially trouble-free at current rated power until the current 1985 outage. Strategies and lessons learned that contributed to a successful RTS (e.g., all regulatory and economic constraints were met) included the use of numerous brainstorm sessions between SCE, Bechtel, and other consultants; innovative state of the art technical criterion (e.g., realistic versus overly conservative engineering); and budget and schedule contingencies consistent with the known and unknown risks

Description of the structural evolution of a hydrating portland cement paste by SANS

International Nuclear Information System (INIS)

Haeussler, F.; Eichhorn, F.; Baumbach, H.

1994-01-01

On the spectrometer MURN at the pulsed reactor IBR-2 dry Portland cement, silica fume, and a hydrating Portland cement paste were studied by small-angle neutron scattering (SANS). By using the TOF-method a momentum transfer from 0.07 nm -1 to 7 nm -1 is detectable. Every component (dry cement powder, clinker minerals, hydrating cement pastes) shows a different scattering behaviour. In the measured Q-region the hardening cement paste does not show a Porod-like behaviour of SANS-curves. In contrast the Porod's potential law holds for dry powder samples of clinker minerals and silica fume. In experiments carried out to observe the hydration progress within the first 321 days the characteristics of the scattering curves (potential behaviour, the radius of gyration, and the macroscopic scattering cross section at Q = 0 nm -1 were measured. Some evolution of the inner structure of the hardened cement paste was noted. (orig.)

76 FR 10945 - San Luis Trust Bank, FSB, San Luis Obispo, CA; Notice of Appointment of Receiver

Science.gov (United States)

2011-02-28

... DEPARTMENT OF THE TREASURY Office of Thrift Supervision San Luis Trust Bank, FSB, San Luis Obispo, CA; Notice of Appointment of Receiver Notice is hereby given that, pursuant to the authority... appointed the Federal Deposit Insurance Corporation as sole Receiver for San Luis Trust Bank, FSB, San Luis...

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

Le, T.T.; Pevey, R.E.

1993-01-01

The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

76 FR 22809 - Safety Zone; Bay Ferry II Maritime Security Exercise; San Francisco Bay, San Francisco, CA

Science.gov (United States)

2011-04-25

... DEPARTMENT OF HOMELAND SECURITY Coast Guard 33 CFR Part 165 [Docket No. USCG-2011-0196] RIN 1625-AA00 Safety Zone; Bay Ferry II Maritime Security Exercise; San Francisco Bay, San Francisco, CA AGENCY... Security Exercise; San Francisco Bay, San Francisco, CA. (a) Location. The limits of this safety zone...

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

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

Assessment of Radiation Embrittlement in Nuclear Reactor Pressure Vessel Surrogate Materials

Science.gov (United States)

Balzar, Davor

2010-10-01

The radiation-enhanced formation of small (1-2 nm) copper-rich precipitates (CRPs) is critical for the occurrence of embrittlement in nuclear-reactor pressure vessels. Small CRPs are coherent with the bcc matrix, which causes local matrix strain and interaction with the dislocation strain fields, thus impeding dislocation mobility. As CRPs grow, there is a critical size at which a phase transformation occurs, whereby the CRPs are no longer coherent with the matrix, and the strain is relieved. Diffraction-line-broadening analysis (DLBA) and small-angle neutron scattering (SANS) were used to characterize the precipitate formation in surrogate ferritic reactor-pressure vessel steels. The materials were aged for different times at elevated temperature to produce a series of specimens with different degrees of copper precipitation. SANS measurements showed that the precipitate size distribution broadens and shifts toward larger sizes as a function of ageing time. Mechanical hardness showed an increase with ageing time, followed by a decrease, which can be associated with the reduction in the number density as well as the loss of coherency at larger sizes. Inhomogeneous strain correlated with mechanical hardness.

The San Bernabe power substation; La subestacion San Bernabe

Energy Technology Data Exchange (ETDEWEB)

Chavez Sanudo, Andres D. [Luz y Fuerza del Centro, Mexico, D. F. (Mexico)

1997-12-31

The first planning studies that gave rise to the San Bernabe substation go back to year 1985. The main circumstance that supports this decision is the gradual restriction for electric power generation that has been suffering the Miguel Aleman Hydro System, until its complete disappearance, to give priority to the potable water supply through the Cutzamala pumping system, that feeds in an important way Mexico City and the State of Mexico. In this document the author describes the construction project of the San Bernabe Substation; mention is made of the technological experiences obtained during the construction and its geographical location is shown, as well as the one line diagram of the same [Espanol] Los primeros estudios de planeacion que dieron origen a la subestacion San Bernabe se remontan al ano de 1985. La circunstancia principal que soporta esta decision es la restriccion paulatina para generar energia que ha venido experimentando el Sistema Hidroelectrico Miguel Aleman, hasta su desaparicion total, para dar prioridad al suministro de agua potable por medio del sistema de bombeo Cutzamala, que alimenta en forma importante a la Ciudad de Mexico y al Estado de Mexico. En este documento el autor describe el proyecto de construccion de la subestacion San Bernabe; se mencionan las experiencias tecnologicas obtenidas durante su construccion y se ilustra su ubicacion geografica, asi como un diagrama unifilar de la misma

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

International Nuclear Information System (INIS)

Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

2012-01-01

10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

April 1906 San Francisco, USA Images

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The 1906 San Francisco earthquake was the largest event (magnitude 8.3) to occur in the conterminous United States in the 20th Century. Recent estimates indicate...

The life-extension and upgrade program of the Tsing Hua Open-pool Reactor (THOR) and its research prospectives

International Nuclear Information System (INIS)

Kai, J.-J.

1992-01-01

The Tsing Hua Open-Pool Reactor (THOR) has been operated for thirty years. It is the regulations of the ROCAEC that any reactor shall be decommissioned after forty-year operation since the first fuel loading. Therefore, for extending the lifetime of THOR, it is necessary to have a life-extension program to be approved by the ROCAEC and also completed by the year of 1997. At the same time, for proceeding new research purposes, it is planed to upgrade the thermal power of THOR from 1 Wth up to 3 Wth and hopefully to reach the maximum thermal neutron flux of 5x10 13 n/cm 2 .s and the fast flux close to that order. New research directions involve (a) boron-captured neutron cancer therapy (BNCT) (b) small-angle neutron scattering (SANS). (author)

Description of gravity cores from San Pablo Bay and Carquinez Strait, San Francisco Bay, California

Science.gov (United States)

Woodrow, Donald L.; John L. Chin,; Wong, Florence L.; Fregoso, Theresa A.; Jaffe, Bruce E.

2017-06-27

Seventy-two gravity cores were collected by the U.S. Geological Survey in 1990, 1991, and 2000 from San Pablo Bay and Carquinez Strait, California. The gravity cores collected within San Pablo Bay contain bioturbated laminated silts and sandy clays, whole and broken bivalve shells (mostly mussels), fossil tube structures, and fine-grained plant or wood fragments. Gravity cores from the channel wall of Carquinez Strait east of San Pablo Bay consist of sand and clay layers, whole and broken bivalve shells (less than in San Pablo Bay), trace fossil tubes, and minute fragments of plant material.

78 FR 34123 - Notice of Inventory Completion: San Francisco State University NAGPRA Program, San Francisco, CA

Science.gov (United States)

2013-06-06

... completion of an inventory of human remains and associated funerary objects under the control of the San....R50000] Notice of Inventory Completion: San Francisco State University NAGPRA Program, San Francisco, CA... NAGPRA Program has completed an inventory of human remains and associated funerary objects, in...

78 FR 21403 - Notice of Inventory Completion: San Francisco State University NAGPRA Program, San Francisco, CA

Science.gov (United States)

2013-04-10

... completion of an inventory of human remains and associated funerary objects under the control of the San....R50000] Notice of Inventory Completion: San Francisco State University NAGPRA Program, San Francisco, CA... NAGPRA Program has completed an inventory of human remains and associated funerary objects, in...

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

International Nuclear Information System (INIS)

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

1992-01-01

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

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

Hollis, A.A.; Mitchell, J.T.D.

1977-12-01

This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

75 FR 15611 - Safety Zone; United Portuguese SES Centennial Festa, San Diego Bay, San Diego, CA

Science.gov (United States)

2010-03-30

...-AA00 Safety Zone; United Portuguese SES Centennial Festa, San Diego Bay, San Diego, CA AGENCY: Coast... navigable waters of the San Diego Bay in support of the United Portuguese SES Centennial Festa. This... Centennial Festa, which will include a fireworks presentation originating from a tug and barge combination in...

Submerged anaerobic membrane bioreactor (SAnMBR) performance on sewage treatment: removal efficiencies, biogas production and membrane fouling.

Science.gov (United States)

Chen, Rong; Nie, Yulun; Ji, Jiayuan; Utashiro, Tetsuya; Li, Qian; Komori, Daisuke; Li, Yu-You

2017-09-01

A submerged anaerobic membrane reactor (SAnMBR) was employed for comprehensive evaluation of sewage treatment at 25 °C and its performance in removal efficiency, biogas production and membrane fouling. Average 89% methanogenic degradation efficiency as well as 90%, 94% and 96% removal of total chemical oxygen demand (TCOD), biochemical oxygen demand (BOD) and nonionic surfactant were obtained, while nitrogen and phosphorus were only subjected to small removals. Results suggest that SAnMBRs can effectively decouple organic degradation and nutrients disposal, and reserve all the nitrogen and phosphorus in the effluent for further possible recovery. Small biomass yields of 0.11 g mixed liquor volatile suspended solids (MLVSS)/gCOD were achieved, coupled to excellent methane production efficiencies of 0.338 NLCH 4 /gCOD, making SAnMBR an attractive technology characterized by low excess sludge production and high bioenergy recovery. Batch tests revealed the SAnMBR appeared to have the potential to bear a high food-to-microorganism ratio (F/M) of 1.54 gCOD/gMLVSS without any inhibition effect, and maximum methane production rate occurred at F/M 0.7 gCOD/gMLVSS. Pore blocking dominated the membrane fouling behaviour at a relative long hydraulic retention time (HRT), i.e. >12 hours, while cake layer dominated significantly at shorter HRTs, i.e. <8 hours.

Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grosse, M; Boehmert, J; Gilles, R [Hahn-Meitner-Institut Berlin GmbH (Germany)

1998-10-01

The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

International Nuclear Information System (INIS)

Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

2014-01-01

The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

Performance of BATAN-SANS instrument

Energy Technology Data Exchange (ETDEWEB)

Ikram, Abarrul; Insani, Andon [National Nuclear Energy Agency, P and D Centre for Materials Science and Technology, Serpong (Indonesia)

2003-03-01

SANS data from some standard samples have been obtained using BATAN-SANS instrument in Serpong. The experiments were performed for various experimental set-ups that involve different detector positions and collimator lengths. This paper describes the BATAN-SANS instrument briefly as well as the data taken from those experiments and followed with discussion of the results concerning the performance and calibration of the instrument. The standard samples utilized in these experiments include porous silica, polystyrene-poly isoprene, silver behenate, poly ball and polystyrene-poly (ethylene-alt-propylene). Even though the results show that BATAN-SANS instrument is in good shape, but rooms for improvements are still widely open especially for the velocity selector and its control system. (author)

The World's Reactors no. 70 - Forsmark 3, BWR-75

International Nuclear Information System (INIS)

Anon.

1976-01-01

A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

A case for historic joint rupture of the San Andreas and San Jacinto faults

OpenAIRE

Lozos, Julian C.

2016-01-01

The San Andreas fault is considered to be the primary plate boundary fault in southern California and the most likely fault to produce a major earthquake. I use dynamic rupture modeling to show that the San Jacinto fault is capable of rupturing along with the San Andreas in a single earthquake, and interpret these results along with existing paleoseismic data and historic damage reports to suggest that this has likely occurred in the historic past. In particular, I find that paleoseismic data...

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

Sediment transport of streams tributary to San Francisco, San Pablo, and Suisun Bays, California, 1909-66

Science.gov (United States)

Porterfield, George

1980-01-01

A review of historical sedimentation data is presented, results of sediment-data collection for water years 1957-59 are summarized, and long-term sediment-discharge estimates from a preliminary report are updated. Comparison of results based on 3 years of data to those for the 10 water years, 1957-66, provides an indication of the adequacy of the data obtained during the short period to define the long-term relation between sediment transport and streamflow. During 1909-66, sediment was transported to the entire San Francisco Bay system at an average rate of 8.6 million cubic yards per year. The Sacramento and San Joaquin River basins provided about 83% of the sediment inflow to the system annually during 1957-66 and 86% during 1909-66. About 98% of this inflow was measured or estimated at sediment measuring sites. Measured sediment inflow directly to the bays comprised only about 40% of the total discharged by basins directly tributary to the bays. About 90% of the total sediment discharge to the delta and the bays in the San Francisco Bay system thus was determined on the basis of systematic measurements. (USGS)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-11-01

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

Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

Holocene slip rates along the San Andreas Fault System in the San Gorgonio Pass and implications for large earthquakes in southern California

Science.gov (United States)

Heermance, Richard V.; Yule, Doug

2017-06-01

The San Gorgonio Pass (SGP) in southern California contains a 40 km long region of structural complexity where the San Andreas Fault (SAF) bifurcates into a series of oblique-slip faults with unknown slip history. We combine new 10Be exposure ages (Qt4: 8600 (+2100, -2200) and Qt3: 5700 (+1400, -1900) years B.P.) and a radiocarbon age (1260 ± 60 years B.P.) from late Holocene terraces with scarp displacement of these surfaces to document a Holocene slip rate of 5.7 (+2.7, -1.5) mm/yr combined across two faults. Our preferred slip rate is 37-49% of the average slip rates along the SAF outside the SGP (i.e., Coachella Valley and San Bernardino sections) and implies that strain is transferred off the SAF in this area. Earthquakes here most likely occur in very large, throughgoing SAF events at a lower recurrence than elsewhere on the SAF, so that only approximately one third of SAF ruptures penetrate or originate in the pass.Plain Language SummaryHow large are earthquakes on the southern San Andreas Fault? The answer to this question depends on whether or not the earthquake is contained only along individual fault sections, such as the Coachella Valley section north of Palm Springs, or the rupture crosses multiple sections including the area through the San Gorgonio Pass. We have determined the age and offset of faulted stream deposits within the San Gorgonio Pass to document slip rates of these faults over the last 10,000 years. Our results indicate a long-term slip rate of 6 mm/yr, which is almost 1/2 of the rates east and west of this area. These new rates, combined with faulted geomorphic surfaces, imply that large magnitude earthquakes must occasionally rupture a 300 km length of the San Andreas Fault from the Salton Sea to the Mojave Desert. Although many ( 65%) earthquakes along the southern San Andreas Fault likely do not rupture through the pass, our new results suggest that large >Mw 7.5 earthquakes are possible on the southern San Andreas Fault and likely

New fault picture points toward San Francisco Bay area earthquakes

Science.gov (United States)

Kerr, R. A.

1989-01-01

Recent earthquakes and a new way of looking at faults suggest that damaging earthquakes are closing in on the San Francisco area. Earthquakes Awareness Week 1989 in northern California started off with a bang on Monday, 3 April, when a magnitude 4.8 earthquake struck 15 kilometers northeast of San Jose. The relatively small shock-its primary damage was the shattering of an air-control tower window-got the immediate attention of three U.S Geological Survey seismologists in Menlo Park near San Francisco. David Oppenheimer, William Bakun, and Allan Lindh had forecast a nearby earthquake in a just completed report, and this, they thought, might be it. 

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

77 FR 34988 - Notice of Inventory Completion: San Diego State University, San Diego, CA

Science.gov (United States)

2012-06-12

.... ACTION: Notice. SUMMARY: San Diego State University Archeology Collections Management Program has... that believes itself to be culturally affiliated with the human remains and associated funerary objects may contact San Diego State University Archeology Collections Management Program. Repatriation of the...

3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

A robot-automated work site for repair of the Chinon A3 reactor

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

International Nuclear Information System (INIS)

Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

1987-01-01

In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

Decommissioning of the BR3 pressurized-water reactor

International Nuclear Information System (INIS)

Massaut, V.

1996-01-01

The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

TRIGA 14 MW Research Reactor Status and Utilization

International Nuclear Information System (INIS)

Barbos, D.; Ciocanescu, M.; Paunoiu, C.

2016-01-01

neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector, mainly for fission and fusion. At the TRIGA 14 MW reactor a neutron diffractometer and a SANS device are available for material residual stress and texture measurement. (author)

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

LaRC Modeling of Ozone Formation in San Antonio, Texas

Science.gov (United States)

Guo, F.; Griffin, R. J.; Bui, A.; Schulze, B.; Wallace, H. W., IV; Flynn, J. H., III; Erickson, M.; Kotsakis, A.; Alvarez, S. L.; Usenko, S.; Sheesley, R. J.; Yoon, S.

2017-12-01

Ozone (O3) is one of the most important trace species within the troposphere and results from photochemistry involving emissions from a complex array of sources. Ground-level O3 is detrimental to ecosystems and causes a variety of human health problems including respiratory irritation, asthma and reduction in lung capacity. However, the O3 Design Value in San Antonio, Texas, was in violation of the federal threshold set by the EPA (70 ppb, 8-hr max) based on the average for the most recent three-year period (2014-2016). To understand the sources of high O3 concentrations in this nonattainment area, we assembled and deployed a mobile air quality laboratory and operated it in two locations in the southeast (Traveler's World RV Park) and northwest (University of Texas at San Antonio) of downtown San Antonio during summer 2017 to measure O3 and its precursors, including total nitrogen oxides (NOx) and volatile organic compounds (VOCs). Additional measurements included temperature, relative humidity, pressure, solar radiation, wind speed, wind direction, total reactive nitrogen (NOy), carbon monoxide (CO), and aerosol composition and concentration. We will use the campaign data and the NASA Langley Research Center (LaRC) Zero-Dimensional Box Model (Crawford et al., 1999; Olson et al., 2006) to calculate O3 production rate, NOx and hydroxyl radical chain length, and NOx versus VOCs sensitivity at different times of a day with different photochemical and meteorological conditions. A key to our understanding is to combine model results with measurements of precursor gases, particle chemistry and particle size to support the identification of O3 sources, its major formation pathways, and how the ozone production efficiency (OPE) depends on various factors. The resulting understanding of the causes of high O3 concentrations in the San Antonio area will provide insight into future air quality protection.

Effectiveness of Kampo medicine Gorei-san for chronic subdural hematoma

International Nuclear Information System (INIS)

Miyagami, Mitsusuke; Kagawa, Yukihide

2009-01-01

Chronic subdural hematomas (CSDHs) are basically treated by surgery. In some cases with no or minimum symptoms, however, they may be treated conservatively. In the present study, we evaluated the therapeutic effect of a Kampo medicine (Japanese traditional herbal medicine), Gorei-san, in the treatment of those CSDHs. Gorei-san 7.5 g t.i.d. was orally administered for 4 weeks in 22 patients with 27 CSDHs. Maximum thickness of the hematoma was followed up on CT scan for 4 to 29 weeks after administration of Gorei-san. In 7 of 22 patients, tranexamic acid and/or carbazochrome sodium sulfonate were also administrated. Gorei-san was effective in 23 of 27 CSDHs. In 12 of them, the hematoma was completely disappeared within 14 weeks after administration. In the other 11 CSDHs, the thickness was decreased. In those effective cases, thickness began to decrease 3 to 4 weeks after administration of Gorei-san. It was more effective in CSDHs with iso-/high or mixed density than with low density on CT. It was not effective in 4 out of 27 CSDHs. No apparent adverse effect was noted in the present series of patients. The present study suggests that a Kampo medicine, Gorei-san, is a useful option in the conservative treatment of CSDHs with no or minimum symptoms. (author)

Operational experience with PWR secondary water chemistry: a panel presentation San Onofre Unit 1

International Nuclear Information System (INIS)

Britt, R.D.; Millard, R.E.; DiFilippo, M.N.

1975-01-01

The three steam generators have been on phosphate chemistry since startup except for one brief period when volatile chemistry was attempted. Initially, coordinated pH-phosphate control was recommended by Westinghouse for the steam generators; however, after one year of operation, Westinghouse recommended changing to congruent control. From startup in 1967 until the end of 1970, the Na/PO 4 molar ratio was generally maintained in the 2.6 to 2.8 range, with a 5 to 10 ppM phosphate residual. A summary of steam generator chemistry from initial startup to the present is presented

New evidence on the state of stress of the san andreas fault system.

Science.gov (United States)

Zoback, M D; Zoback, M L; Mount, V S; Suppe, J; Eaton, J P; Healy, J H; Oppenheimer, D; Reasenberg, P; Jones, L; Raleigh, C B; Wong, I G; Scotti, O; Wentworth, C

1987-11-20

Contemporary in situ tectonic stress indicators along the San Andreas fault system in central California show northeast-directed horizontal compression that is nearly perpendicular to the strike of the fault. Such compression explains recent uplift of the Coast Ranges and the numerous active reverse faults and folds that trend nearly parallel to the San Andreas and that are otherwise unexplainable in terms of strike-slip deformation. Fault-normal crustal compression in central California is proposed to result from the extremely low shear strength of the San Andreas and the slightly convergent relative motion between the Pacific and North American plates. Preliminary in situ stress data from the Cajon Pass scientific drill hole (located 3.6 kilometers northeast of the San Andreas in southern California near San Bernardino, California) are also consistent with a weak fault, as they show no right-lateral shear stress at approximately 2-kilometer depth on planes parallel to the San Andreas fault.

Lipid based drug delivery systems: Kinetics by SANS

Science.gov (United States)

Uhríková, D.; Teixeira, J.; Hubčík, L.; Búcsi, A.; Kondela, T.; Murugova, T.; Ivankov, O. I.

2017-05-01

N,N-dimethyldodecylamine-N-oxide (C12NO) is a surfactant that may exist either in a neutral or protonated form depending on the pH of aqueous solutions. Using small angle X-ray diffraction (SAXD) we demonstrate structural responsivity of C12NO/dioleoylphospha-tidylethanolamine (DOPE)/DNA complexes designed as pH sensitive gene delivery vectors. Small angle neutron scattering (SANS) was employed to follow kinetics of C12NO protonization and DNA binding into C12NO/DOPE/DNA complexes in solution of 150 mM NaCl at acidic condition. SANS data analyzed using paracrystal lamellar model show the formation of complexes with stacking up to ∼32 bilayers, spacing ∼ 62 Å, and lipid bilayer thickness ∼37 Å in 3 minutes after changing pH from 7 to 4. Subsequent structural reorganization of the complexes was observed along 90 minutes of SANS mesurements.

Lipid based drug delivery systems: Kinetics by SANS

International Nuclear Information System (INIS)

Uhríková, D; Hubčík, L; Búcsi, A; Kondela, T; Teixeira, J; Murugova, T; Ivankov, O I

2017-01-01

N,N-dimethyldodecylamine-N-oxide (C 12 NO) is a surfactant that may exist either in a neutral or protonated form depending on the pH of aqueous solutions. Using small angle X-ray diffraction (SAXD) we demonstrate structural responsivity of C 12 NO/dioleoylphospha-tidylethanolamine (DOPE)/DNA complexes designed as pH sensitive gene delivery vectors. Small angle neutron scattering (SANS) was employed to follow kinetics of C 12 NO protonization and DNA binding into C 12 NO/DOPE/DNA complexes in solution of 150 mM NaCl at acidic condition. SANS data analyzed using paracrystal lamellar model show the formation of complexes with stacking up to ∼32 bilayers, spacing ∼ 62 Å, and lipid bilayer thickness ∼37 Å in 3 minutes after changing pH from 7 to 4. Subsequent structural reorganization of the complexes was observed along 90 minutes of SANS mesurements. (paper)

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.)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Advances in Reactor physics, mathematics and computation. Volume 3

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

International Nuclear Information System (INIS)

Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

2004-01-01

This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5 percent respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock

Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

International Nuclear Information System (INIS)

Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

2004-01-01

This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation--or neutrino oscillation--by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5% respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical

Tratamiento de las excretas de cerdo mediante un reactor anaeróbico SCFBR a nivel de banco Treatment of pig excreta using an SCFBR anaerobic reactor

Directory of Open Access Journals (Sweden)

Caicedo Luis A.

1999-06-01

Full Text Available sans-serif";; mso-fareast-font-family: Calibri; mso-fareast-theme-font: minor-latin; mso-ansi-language: ES-CO; mso-fareast-language: EN-US; mso-bidi-language: AR-SA;" lang="ES-CO">Un nuevo reactor anaeróbico denominado Sludge Central Fixed Bed Reactor (SCFBR fue construido y evaluado para tratar los residuos líquidos de las granjas porcícolas. El SCFBR está constituido por tres zonas principales. Una zona inferior de lodos, seguida por un módulo empacado ubicado en forma concéntrica y, en la parte superior, una zona de separación sólido, líquido y gas. El reactor de 28,5 1 de volumen de reacción fue evaluado durante 210 días para tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/ 1 día. El SCFBR fue alimentado inicialmente en forma discontinua con tiempos de retención hidráulicos (TRH de 10 y 10,7 días. Posteriormente el TRH fue disminuido a 3,87 días con una alimentación en continuo. Para las tres cargas orgánicas de 0,548, 0,421 y 1,239 g DQO/1 día se obtuvieron remociones en la demanda química de oxígeno (DQO de 68%, 81% y 73% y en los sólidos volátiles (SV, de 53,5%, 55,8% y 50,1%, respectivamente. El SCFBR presentó un buen desempeño, re-presentado en las eficiencias de remoción y en la estabilidad observada. Se presenta una microfotografía tomada de una muestra de lodo de la zona inferior del SCFBR, observándose una gran presencia de microorganismos del género Methanosaeta (Methanothrix.

sans-serif";; mso-ansi-language: EN-US;" lang="EN-US">A new anaerobic reactor called the Sludge Central Fixed Bed Reactor (SCFBR was built and evaluated for the treatment of liquid residue from the pig farms. The SCFBR has three main parts. The lower area is for sludge, the middle part consists of a concentrically

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

Technical specifications: San Onofre Nuclear Generating Station, Unit No. 3, Docket No. 50-362. Appendix A to license No. NPF-15

International Nuclear Information System (INIS)

1982-11-01

This report includes the following sections: definitions, safety limits and limiting safety system settings, limiting conditions for operation and surveillance requirements, bases, design features, and administrative controls

78 FR 53243 - Safety Zone; TriRock San Diego, San Diego Bay, San Diego, CA

Science.gov (United States)

2013-08-29

... this rule because the logistical details of the San Diego Bay triathlon swim were not finalized nor... September 22, 2013. (c) Definitions. The following definition applies to this section: Designated...

Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.

1994-01-01

TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

On the major DYN3D developments for fast reactor design and transient analysis

International Nuclear Information System (INIS)

Merk, B.; Kliem, S.

2013-01-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

On the major DYN3D developments for fast reactor design and transient analysis

Energy Technology Data Exchange (ETDEWEB)

Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

2013-07-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Structure of the 1906 near-surface rupture zone of the San Andreas Fault, San Francisco Peninsula segment, near Woodside, California

Science.gov (United States)

Rosa, C.M.; Catchings, R.D.; Rymer, M.J.; Grove, Karen; Goldman, M.R.

2016-07-08

High-resolution seismic-reflection and refraction images of the 1906 surface rupture zone of the San Andreas Fault near Woodside, California reveal evidence for one or more additional near-surface (within about 3 meters [m] depth) fault strands within about 25 m of the 1906 surface rupture. The 1906 surface rupture above the groundwater table (vadose zone) has been observed in paleoseismic trenches that coincide with our seismic profile and is seismically characterized by a discrete zone of low P-wave velocities (Vp), low S-wave velocities (Vs), high Vp/Vs ratios, and high Poisson’s ratios. A second near-surface fault strand, located about 17 m to the southwest of the 1906 surface rupture, is inferred by similar seismic anomalies. Between these two near-surface fault strands and below 5 m depth, we observed a near-vertical fault strand characterized by a zone of high Vp, low Vs, high Vp/Vs ratios, and high Poisson’s ratios on refraction tomography images and near-vertical diffractions on seismic-reflection images. This prominent subsurface zone of seismic anomalies is laterally offset from the 1906 surface rupture by about 8 m and likely represents the active main (long-term) strand of the San Andreas Fault at 5 to 10 m depth. Geometries of the near-surface and subsurface (about 5 to 10 m depth) fault zone suggest that the 1906 surface rupture dips southwestward to join the main strand of the San Andreas Fault at about 5 to 10 m below the surface. The 1906 surface rupture forms a prominent groundwater barrier in the upper 3 to 5 m, but our interpreted secondary near-surface fault strand to the southwest forms a weaker barrier, suggesting that there has been less or less-recent near-surface slip on that strand. At about 6 m depth, the main strand of the San Andreas Fault consists of water-saturated blue clay (collected from a hand-augered borehole), which is similar to deeply weathered serpentinite observed within the main strand of the San Andreas Fault at

RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija 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 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

77 FR 59969 - Notice of Inventory Completion: San Francisco State University, Department of Anthropology, San...

Science.gov (United States)

2012-10-01

... Inventory Completion: San Francisco State University, Department of Anthropology, San Francisco, CA... Francisco State University, NAGPRA Program (formerly in the Department of Anthropology). The human remains... State University Department of Anthropology records. In the Federal Register (73 FR 30156-30158, May 23...

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

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

Study of seismic responses of Candu-3 reactor building using isolator bearings

International Nuclear Information System (INIS)

Biswas, J.K.

1992-01-01

Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

33 CFR 165.1182 - Safety/Security Zone: San Francisco Bay, San Pablo Bay, Carquinez Strait, and Suisun Bay, CA.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety/Security Zone: San... Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) PORTS AND WATERWAYS SAFETY... Areas Eleventh Coast Guard District § 165.1182 Safety/Security Zone: San Francisco Bay, San Pablo Bay...

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

An expert system technology for work authorization information systems

International Nuclear Information System (INIS)

Munchausen, J.H.; Glazer, K.A.

1988-01-01

This paper describes the effort by Southern California Edison Company (SCE) and the Electric Power Research Institute (EPRI) to develop an expert systems work station designed to support the San Onofre Nuclear Generating Station (SONGS). The expert systems work station utilizes IntelliCorp KEE (Knowledge Engineering Environment) and EPRI-IntelliCorp PLEXSYS (PLant EXpert SYStem) technology, and SCE Piping and Instrumentation Diagrams (P and ID's) and host-based computer applications to assist plant operations and maintenance personnel in the development of safety tagout boundaries. Of significance in this venture is the merging of conventional computer applications technology with expert systems technology. The EPRI PLEXSYS work station will act as a front-end for the SONGS Tagout Administration and Generation System (TAGS), a conventional CICS/COBOL mainframe computer application

SAN MICHELE. ENTRE CIELO Y MAR / San Michele, between sky and sea

Directory of Open Access Journals (Sweden)

Pablo Blázquez Jesús

2012-11-01

Full Text Available RESUMEN El cementerio es uno de los tipos arquitectónicos más profundos y metafóricos. El concurso para la ampliación del cementerio de San Michele, convocado en 1998 por la administración Municipal de Venecia, se convierte en un excelente campo de pruebas sobre el que poder analizar el contexto histórico en torno a esta tipología, y su relación con la ciudad y el territorio. El estudio de este caso concreto nos permite descubrir personajes, relaciones casuales y hallazgos que se despliegan a lo largo del texto. La historia del cementerio de San Michele es también la crónica de la transformación de la ciudad de Venecia y su Laguna. Interpretando este concurso como un instrumento de investigación, el objetivo del artículo es el de comprender la realidad contemporánea de la arquitectura funeraria a través de la isla de San Michele, Venecia, y las propuestas finalistas de Carlos Ferrater, Enric Miralles y David Chipperfield. Una historia bajo la cual se vislumbran claves que nos sirven para reflexionar acerca del cementerio contemporáneo, la ciudad y el territorio. SUMMARY The cemetery is one of the most profound and metaphorical kinds of architecture. The competition for the extension of the San Michele Cemetery, called in 1998 by the Venice municipal administration, is an excellent testing ground on which to analyse the historical context surrounding this type of architecture, and its relationship with the city and the region. The study of this particular case allows us to uncover characters, casual relationships and findings that unfold throughout the text. The history of the San Michele cemetery is also the chronicle of the transformation of the city of Venice and its Lagoon. Interpreting this competition as a research tool, the aim of the paper is to understand the contemporary reality of funerary architecture through the island of San Michele, Venice, and the finalist proposals of Carlos Ferrater, Enric Miralles and David

Pleistocene Brawley and Ocotillo Formations: Evidence for initial strike-slip deformation along the San Felipe and San Jacinto fault zonez, Southern California

Science.gov (United States)

Kirby, S.M.; Janecke, S.U.; Dorsey, R.J.; Housen, B.A.; Langenheim, V.E.; McDougall, K.A.; Steeley, A.N.

2007-01-01

We examine the Pleistocene tectonic reorganization of the Pacific-North American plate boundary in the Salton Trough of southern California with an integrated approach that includes basin analysis, magnetostratigraphy, and geologic mapping of upper Pliocene to Pleistocene sedimentary rocks in the San Felipe Hills. These deposits preserve the earliest sedimentary record of movement on the San Felipe and San Jacinto fault zones that replaced and deactivated the late Cenozoic West Salton detachment fault. Sandstone and mudstone of the Brawley Formation accumulated between ???1.1 and ???0.6-0.5 Ma in a delta on the margin of an arid Pleistocene lake, which received sediment from alluvial fans of the Ocotillo Formation to the west-southwest. Our analysis indicates that the Ocotillo and Brawley formations prograded abruptly to the east-northeast across a former mud-dominated perennial lake (Borrego Formation) at ???1.1 Ma in response to initiation of the dextral-oblique San Felipe fault zone. The ???25-km-long San Felipe anticline initiated at about the same time and produced an intrabasinal basement-cored high within the San Felipe-Borrego basin that is recorded by progressive unconformities on its north and south limbs. A disconformity at the base of the Brawley Formation in the eastern San Felipe Hills probably records initiation and early blind slip at the southeast tip of the Clark strand of the San Jacinto fault zone. Our data are consistent with abrupt and nearly synchronous inception of the San Jacinto and San Felipe fault zones southwest of the southern San Andreas fault in the early Pleistocene during a pronounced southwestward broadening of the San Andreas fault zone. The current contractional geometry of the San Jacinto fault zone developed after ???0.5-0.6 Ma during a second, less significant change in structural style. ?? 2007 by The University of Chicago. All rights reserved.

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

Research on burnup physics

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1974-07-01

One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Solar-energy system performance evaluation. San Anselmo School, San Jose, California, July 1980-March 1981

Energy Technology Data Exchange (ETDEWEB)

Pakkala, P.A.

1981-01-01

The San Anselmo School is a one-story, brick elementary school building located in San Jose, California. The active solar energy system is designed to supply 70% of the heating load and 72% of the cooling load. It is equipped with 3.740 square feet of evacuated tube collectors, 2175-gallon tank for storage, four auxiliary gas-fired absorption chiller/heaters, and a solar-supplied absorption chiller. The measured heating and cooling solar fractions were 9% and 19%, respectively, for an overall solar fraction of 16%, the lowered performance being attributed to severe system control problems. Performance data include the solar savings ratio, conventional fuel savings, system performance factor, and solar system coefficient of performance. Performance data are presented for the overall system and for each subsystem. System operation and solar energy utilization data are included. Also included are a description of the system, performance evaluation techniques, sensor technology, and typical performance data for a month. Weather data are also tabulated. (LEW)

Hydrology of the middle San Pedro area, southeastern Arizona

Science.gov (United States)

Cordova, Jeffrey T.; Dickinson, Jesse; Beisner, Kimberly R.; Hopkins, Candice B.; Kennedy, Jeffrey R.; Pool, Donald R.; Glenn, Edward P.; Nagler, Pamela L.; Thomas, Blakemore E.

2015-05-05

In the middle San Pedro Watershed in southeastern Arizona, groundwater is the primary source of water supply for municipal, domestic, industrial, and agricultural use. The watershed comprises two smaller subareas, the Benson subarea and the Narrows-Redington subarea. Early 21st century projections for heavy population growth in the watershed have not yet become a reality, but increased groundwater withdrawals could have undesired consequences - such as decreased base flow to the San Pedro River, and groundwater-level declines - that would lead to the need to deepen existing wells. This report describes the hydrology, hydrochemistry, water quality, and development of a groundwater budget for the middle San Pedro Watershed, focusing primarily on the elements of groundwater movement that could be most useful for the development of a groundwater modelPrecipitation data from Tombstone, Arizona, and base flow at the stream-gaging station on the San Pedro River at Charleston both show relatively dry periods during the 1960s through the mid-1980s and in the mid-1990s to 2009, and wetter periods from the mid-1980s through the mid-1990s. Water levels in four out of five wells near the mountain fronts show cyclical patterns of recharge, with rates of recharge greatest in the early 1980s through the mid-1990s. Three wells near the San Pedro River recorded their lowest levels during the 1950s to the mid-1960s. The water-level record from one well, completed in the confined part of the coarse-grained lower basin fill, showed a decline of approximately 21 meters.Annual flow of the San Pedro River, measured at the Charleston and Redington gages, has decreased since the 1940s. The median annual streamflow and base flow at the gaging station on the river near Tombstone has decreased by 50 percent between the periods 1968–1986 and 1997–2009. Estimates of streamflow infiltration along the San Pedro River during 1914–2009 have decreased 44 percent, with the largest decreases in

Vegetation - San Felipe Valley [ds172

Data.gov (United States)

California Natural Resource Agency — This Vegetation Map of the San Felipe Valley Wildlife Area in San Diego County, California is based on vegetation samples collected in the field in 2002 and 2005 and...

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

El urbanismo de Santiago de Compostela : un plano con las plazuelas de San Martín y de San Miguel de 1709

Directory of Open Access Journals (Sweden)

Miguel Taín Guzmán

1998-01-01

Full Text Available El presente artículo está dedicado al estudio de un plano inédito de 1709 donde se representan las plazuelas de San Martín y de San Miguel, en el barrio intramuros de la Puerta de la Peña de Santiago de Compostela. Gracias al referido dibujo, analizo al detalle el entramado urbano de ambos espacios públicos y los edificios que los delimitan, particularmente la iglesia de San Martín Pinario, el desaparecido Palacio del Tribunal de la Santa Inquisición y la iglesia parroquial de San Miguel dos Agros.The article focuses on the study of a 1709 inpublished street plan of two squares —San Martín and San Miguel— in the Puerta de la Peña quarter (Santiago de Compostela. This oíd drawing shows the urban framework of both public spaces and also the buildings around: San Martín Pinario, the lost Palacio del Tribunal de la Santa Inquisición and the paroquial church of San Miguel de los Agros.

78 FR 57482 - Safety Zone; America's Cup Aerobatic Box, San Francisco Bay, San Francisco, CA

Science.gov (United States)

2013-09-19

...-AA00 Safety Zone; America's Cup Aerobatic Box, San Francisco Bay, San Francisco, CA AGENCY: Coast Guard... America's Cup air shows. These safety zones are established to provide a clear area on the water for... announced by America's Cup Race Management. ADDRESSES: Documents mentioned in this preamble are part of...

The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

International Nuclear Information System (INIS)

Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

2003-01-01

Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

Technical report on implementation of reactor internal 3D modeling and visual database system

International Nuclear Information System (INIS)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

Technical report on implementation of reactor internal 3D modeling and visual database system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Proof, interpretation and evaluation of radiation-induced microstructural changes in WWER reactor pressure vessel steels

International Nuclear Information System (INIS)

Boehmert, J.; Gokhman, A.; Grosse, M.; Ulbricht, A.

2003-06-01

Neutron embrittlement is a special issue for the VVER-type reactors. One of the fundamentals for a reliable assessment of the current material state is knowledge of the causes and mechanisms of neutron embrittlement. The aim of the project is to understand and to quantify the microstructural appearances due to neutron radiation in VVER-type reactor pressure vessel steels. The material base is a broad variation of irradiation probes taken from the irradiation programme Rheinsberg, surveillance programmes of Russian, Ukrainian or Hungarian NPPs or irradiation experiments with mockup-alloys. The microstructure was investigated by different methods. The small angle neutron scattering (SANS) proved to be the most suitable method. A procedure was developed to determine mean diameter, size distribution and volume fraction of irradiation-induced microstructure from SANS experiments in a reliable and comparable manner. With this method microstructural parameters were systematically determined and the main factors of influence were identified. Apart from the neutron fluence the volume fraction of radiation defects mainly changes with the copper or nickel content whereas phosphorus is hardly relevant. Annealing remedies the radiation-induced microstructural appearances. The ratio between nuclear and magnetic neutron scattering provides information on the type of radiation defects. This leads to the conclusion that the material composition changes the radiation defects. The change occurs gradually rather than abruptly. The radiation defects detected by SANS correlate with the radiation hardening and embrittlement. Generally, the results suggest a bimodal mechanism due to radiation-enhanced and radiation-induced defect evolution. A kinetic model on base of the rate theory approach was established. (orig.)

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

The Demise and Rise of the Coy San

Directory of Open Access Journals (Sweden)

Robert J. Gordon

2014-01-01

Full Text Available Review Article:De Jongh, Michael (2012, Roots and Routes: Karretjie People of the Great Karoo: The Marginalisation of a South African First People, Pretoria: UNISA Press, ISBN 978-1-86888-665-4, 220 pp.Glyn, Patricia (2013, What Dawid Knew: A Journey with the Kruipers, Johannesburg: Picador, ISBN 978-1-77010-304-7, 256 pp.Myburgh, Paul John (2013, The Bushman Winter Has Come: The True Story of the Last Band of /Gwikwe Bushmen on the Great Sand Face, Johannesburg: Penguin, ISBN 978-0-14-353066-4, 234 pp.Taylor, Julie J. (2012, Naming the Land: San Identity and Community Conservation in Namibia’s West Caprivi, Basel: Basler Afrika Bibliographien, ISBN 978-3-905758-25-2, 280 pp.Zips-Mairitsch, Manuela (2013, Lost Lands? (Land Rights of the San in Botswana and the Legal Concept of Indigeneity in Africa, Berlin: Lit Verlag, ISBN 978-3-643-90244-3, 430 pp.

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

77 FR 42649 - Safety Zone: Sea World San Diego Fireworks, Mission Bay; San Diego, CA

Science.gov (United States)

2012-07-20

... 1625-AA00 Safety Zone: Sea World San Diego Fireworks, Mission Bay; San Diego, CA AGENCY: Coast Guard... authorized by the Captain of the Port, or his designated representative. DATES: This rule is effective from 8... to ensure the public's safety. B. Basis and Purpose The Ports and Waterways Safety Act gives the...

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

International Nuclear Information System (INIS)

1992-04-01

Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

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

1992-01-01

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

Annealing of the BR3 reactor pressure vessel

International Nuclear Information System (INIS)

Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

1985-01-01

The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

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

Cataclastic rocks of the San Gabriel fault—an expression of deformation at deeper crustal levels in the San Andreas fault zone

Science.gov (United States)

Anderson, J. Lawford; Osborne, Robert H.; Palmer, Donald F.

1983-10-01

The San Gabriel fault, a deeply eroded late Oligocene to middle Pliocene precursor to the San Andreas, was chosen for petrologic study to provide information regarding intrafault material representative of deeper crustal levels. Cataclastic rocks exposed along the present trace of the San Andreas in this area are exclusively a variety of fault gouge that is essentially a rock flour with a quartz, feldspar, biotite, chlorite, amphibole, epidote, and Fe-Ti oxide mineralogy representing the milled-down equivalent of the original rock (Anderson and Osborne, 1979; Anderson et al., 1980). Likewise, fault gouge and associated breccia are common along the San Gabriel fault, but only where the zone of cataclasis is several tens of meters wide. At several localities, the zone is extremely narrow (several centimeters), and the cataclastic rock type is cataclasite, a dark, aphanitic, and highly comminuted and indurated rock. The cataclastic rocks along the San Gabriel fault exhibit more comminution than that observed for gouge along the San Andreas. The average grain diameter for the San Andreas gouge ranges from 0.01 to 0.06 mm. For the San Gabriel cataclastic rocks, it ranges from 0.0001 to 0.007 mm. Whereas the San Andreas gouge remains particulate to the smallest grain-size, the ultra-fine grain matrix of the San Gabriel cataclasite is composed of a mosaic of equidimensional, interlocking grains. The cataclastic rocks along the San Gabriel fault also show more mineralogiec changes compared to gouge from the San Andreas fault. At the expense of biotite, amphibole, and feldspar, there is some growth of new albite, chlorite, sericite, laumontite, analcime, mordenite (?), and calcite. The highest grade of metamorphism is laumontite-chlorite zone (zeolite facies). Mineral assemblages and constrained uplift rates allow temperature and depth estimates of 200 ± 30° C and 2-5 km, thus suggesting an approximate geothermal gradient of ~50°C/km. Such elevated temperatures imply a

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Master-3.0: multi-purpose analyzer for static and transient effects of reactors

International Nuclear Information System (INIS)

Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

2002-03-01

MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

3 D flow computations under a reactor vessel closure head

International Nuclear Information System (INIS)

Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

1995-12-01

The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

Activation cross sections for the generation of long-lived radionuclides of importance in fusion reactor technology

International Nuclear Information System (INIS)

Pashchenko, A.B.

1993-11-01

The present report contains the Summary of the Second IAEA Research Co-ordination Meeting (RCM) on ''Activation Cross Sections for the Generation of Long-Lived Radionuclides of Importance in Fusion Reactor Technology'' which was hosted by TSI Research at Del Mar near San Diego and held from 29 to 30 April 1993. This RCM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers from TSI Research and Westinghouse Hanford Company. Tables of 14 MeV cross sections and cross sections below 14 MeV are included. The papers prepared and presented by the participants at the meeting has been published as separate report INDC(NDS)-286/L. 3 tabs

The disappearing San of southeastern Africa and their genetic affinities.

Science.gov (United States)

Schlebusch, Carina M; Prins, Frans; Lombard, Marlize; Jakobsson, Mattias; Soodyall, Himla

2016-12-01

Southern Africa was likely exclusively inhabited by San hunter-gatherers before ~2000 years ago. Around that time, East African groups assimilated with local San groups and gave rise to the Khoekhoe herders. Subsequently, Bantu-speaking farmers, arriving from the north (~1800 years ago), assimilated and displaced San and Khoekhoe groups, a process that intensified with the arrival of European colonists ~350 years ago. In contrast to the western parts of southern Africa, where several Khoe-San groups still live today, the eastern parts are largely populated by Bantu speakers and individuals of non-African descent. Only a few scattered groups with oral traditions of Khoe-San ancestry remain. Advances in genetic research open up new ways to understand the population history of southeastern Africa. We investigate the genomic variation of the remaining individuals from two South African groups with oral histories connecting them to eastern San groups, i.e., the San from Lake Chrissie and the Duma San of the uKhahlamba-Drakensberg. Using ~2.2 million genetic markers, combined with comparative published data sets, we show that the Lake Chrissie San have genetic ancestry from both Khoe-San (likely the ||Xegwi San) and Bantu speakers. Specifically, we found that the Lake Chrissie San are closely related to the current southern San groups (i.e., the Karretjie people). Duma San individuals, on the other hand, were genetically similar to southeastern Bantu speakers from South Africa. This study illustrates how genetic tools can be used to assess hypotheses about the ancestry of people who seemingly lost their historic roots, only recalling a vague oral tradition of their origin.

Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

2012-03-15

JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

International Nuclear Information System (INIS)

1993-11-01

In a memo dated 19 August 1993, Secretary O'Leary assigned the Office of Environment, Safety and Health the primary responsibility to identify, characterize, and assess the safety, health, and environmental vulnerabilities of the DOE's existing storage conditions and facilities for the storage of irradiated reactor fuel and other reactor irradiated nuclear materials. This volume is divided into three major sections. Section 1 contains the Working Group Assessment Team reports on the following facilities: Hanford Site, INEL, SRS, Oak Ridge Site, West Valley Site, LANL, BNL, Sandia, General Atomics (San Diego), Babcock ampersand Wilcox (Lynchburg Technology Center), and ANL. Section 2 contains the Vulnerability Development Forms from most of these sites. Section 3 contains the documents used by the Working Group in implementing this initiative

Postseismic relaxation along the San Andreas fault at Parkfield from continuous seismological observations.

Science.gov (United States)

Brenguier, F; Campillo, M; Hadziioannou, C; Shapiro, N M; Nadeau, R M; Larose, E

2008-09-12

Seismic velocity changes and nonvolcanic tremor activity in the Parkfield area in California reveal that large earthquakes induce long-term perturbations of crustal properties in the San Andreas fault zone. The 2003 San Simeon and 2004 Parkfield earthquakes both reduced seismic velocities that were measured from correlations of the ambient seismic noise and induced an increased nonvolcanic tremor activity along the San Andreas fault. After the Parkfield earthquake, velocity reduction and nonvolcanic tremor activity remained elevated for more than 3 years and decayed over time, similarly to afterslip derived from GPS (Global Positioning System) measurements. These observations suggest that the seismic velocity changes are related to co-seismic damage in the shallow layers and to deep co-seismic stress change and postseismic stress relaxation within the San Andreas fault zone.

Examining the life history of an individual from Solcor 3, San Pedro de Atacama: Combining bioarchaeology and archaeological chemistry

International Nuclear Information System (INIS)

Torres-Rouff, Cristina; Knudson, Kelly J

2007-01-01

Detailed life history information using multiple lines of evidence including the identification of geographic origins, health, and body use indicators, can be used to elucidate the complex process of acculturation in the San Pedro de Atacama oases of northern Chile during the Middle Horizon. This paper presents the results of bioarchaeological and archaeological chemical analyses of the skeletal remains of an adult male (tomb 50, catalog number 1948) from the cemetery of Solcor 3 (ca. AD 500-900). Strontium isotope ratios in human tooth enamel reveal information about where a person lived during their childhood, when enamel was being formed. Individual 1948 showed strontium isotope ratios decidedly outside the range of the local San Pedro de Atacama strontium isotope signature. Given these data implying that individual 1948 was originally from elsewhere, an examination of his health status, social role, and mortuary context provides insight into the treatment of foreigners in San Pedro de Atacama. Our data support the argument that individual 1948's foreign birth did not hinder his later assimilation into Atacameno society. He was buried in a local cemetery with a typical mortuary assemblage for a male of this time and no strong evidence of possible foreign origin. Skeletal indicators of diet and activity patterns do not distinguish individual 1948 from the local population, suggesting that his lifestyle was similar to that of other Atacamenos. Therefore, our analyses suggest that individual 1948's acculturation into Atacameno society during his adult life was nearly complete and he retained little to no indication of his probable foreign birth

75 FR 27432 - Security Zone; Golden Guardian 2010 Regional Exercise; San Francisco Bay, San Francisco, CA

Science.gov (United States)

2010-05-17

... can better evaluate its effects on them and participate in the rulemaking process. Small businesses... DEPARTMENT OF HOMELAND SECURITY Coast Guard 33 CFR Part 165 [Docket No. USCG-2010-0221] RIN 1625-AA87 Security Zone; Golden Guardian 2010 Regional Exercise; San Francisco Bay, San Francisco, CA AGENCY...

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Toxic phytoplankton in San Francisco Bay

Science.gov (United States)

Rodgers, Kristine M.; Garrison, David L.; Cloern, James E.

1996-01-01

The Regional Monitoring Program (RMP) was conceived and designed to document the changing distribution and effects of trace substances in San Francisco Bay, with focus on toxic contaminants that have become enriched by human inputs. However, coastal ecosystems like San Francisco Bay also have potential sources of naturally-produced toxic substances that can disrupt food webs and, under extreme circumstances, become threats to public health. The most prevalent source of natural toxins is from blooms of algal species that can synthesize metabolites that are toxic to invertebrates or vertebrates. Although San Francisco Bay is nutrient-rich, it has so far apparently been immune from the epidemic of harmful algal blooms in the world’s nutrient-enriched coastal waters. This absence of acute harmful blooms does not imply that San Francisco Bay has unique features that preclude toxic blooms. No sampling program has been implemented to document the occurrence of toxin-producing algae in San Francisco Bay, so it is difficult to judge the likelihood of such events in the future. This issue is directly relevant to the goals of RMP because harmful species of phytoplankton have the potential to disrupt ecosystem processes that support animal populations, cause severe illness or death in humans, and confound the outcomes of toxicity bioassays such as those included in the RMP. Our purpose here is to utilize existing data on the phytoplankton community of San Francisco Bay to provide a provisional statement about the occurrence, distribution, and potential threats of harmful algae in this Estuary.

Cross-connected onsite emergency A.C. power supplies for multi-unit nuclear power plant sites

International Nuclear Information System (INIS)

Martore, J.A.; Voss, J.D.; Duncil, B.

1987-01-01

Recently, utility management, both at the corporate and plant operations levels, have reinforced their commitment to assuring increased plant reliability and availability. One means of achieving this objective involves an effective preventive maintenance program with technical specifications which allow implementation of certain preventive maintenance without plant shutdown. To accomplish this, Southern California Edison Company (SCE) has proposed a design change for San Onofre nuclear generating station (SONGS) units 2 and 3 to permit on emergency diesel generator for one unit to perform as an available AC power source for both units. Technical specifications for SCE's SONGS units 2 and 3, as at most nuclear power plants, currently require plant shutdown should one of the two dedicated onsite emergency AC power sources (diesel generators) become inoperable for more than 72 hours. This duration hinders root cause failure analysis, tends to limit the flexibility of preventive maintenance and precludes plant operation in the event of component failure. Therefore, this proposed diesel generator cross-connect design change offers an innovative means for averting plant shutdown should a single diesel generator become inoperable for longer than 72 hours. (orig./GL)

Anthropogenic influences on shoreline and nearshore evolution in the San Francisco Bay coastal system

Science.gov (United States)

Dallas, K.L.; Barnard, P.L.

2011-01-01

Analysis of four historical bathymetric surveys over a 132-year period has revealed significant changes to the morphology of the San Francisco Bar, an ebb-tidal delta at the mouth of San Francisco Bay estuary. From 1873 to 2005 the San Francisco Bar vertically-eroded an average of 80 cm over a 125 km2 area, which equates to a total volume loss of 100 ± 52 million m3 of fine- to coarse-grained sand. Comparison of the surveys indicates the entire ebb-tidal delta contracted radially, with the crest moving landward an average of 1 km. Long-term erosion of the ebb-tidal delta is hypothesized to be due to a reduction in the tidal prism of San Francisco Bay and a decrease in coastal sediment supply, both as a result of anthropogenic activities. Prior research indicates that the tidal prism of the estuary was reduced by 9% from filling, diking, and sedimentation. Compilation of historical records dating back to 1900 reveals that a minimum of 200 million m3 of sediment has been permanently removed from the San Francisco Bay coastal system through dredging, aggregate mining, and borrow pit mining. Of this total, ~54 million m3 of sand-sized or coarser sediment was removed from central San Francisco Bay. With grain sizes comparable to the ebb-tidal delta, and its direct connection to the bay mouth, removal of sediments from central San Francisco Bay may limit the sand supply to the delta and open coast beaches. SWAN wave modeling illustrates that changes to the morphology of the San Francisco Bar have altered the alongshore wave energy distribution at adjacent Ocean Beach, and thus may be a significant factor in a persistent beach erosion ‘hot spot’ occurring in the area. Shoreline change analyses show that the sandy shoreline in the shadow of the ebb-tidal delta experienced long-term (1850s/1890s to 2002) and short-term (1960s/1980s to 2002) accretion while the adjacent sandy shoreline exposed to open-ocean waves experienced long-term and short-term erosion. Therefore

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

3D CFD for chemical transport profiles in a rotating disk CVD reactor

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

KAUST Repository

Karsenty, Florent

2012-08-16

Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2015-01-01

Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

SANS study of Th (IV) third phase formation in HNO3 / DHDECMP-n-dodecane system

International Nuclear Information System (INIS)

Lohithakshan, K.V.; Aggarwal, S.K.; Aswal, V.K.

2009-01-01

Third phase formation taking place during the extraction of Th (IV) from nitric acid medium by DHDECMP in dodecane has been investigated by small angle neutron scattering (SANS) and explained the process with Baxter model. (author)

Geologic Map of the San Luis Quadrangle, Costilla County, Colorado

Science.gov (United States)

Machette, Michael N.; Thompson, Ren A.; Drenth, Benjamin J.

2008-01-01

The map area includes San Luis and the primarily rural surrounding area. San Luis, the county seat of Costilla County, is the oldest surviving settlement in Colorado (1851). West of the town are San Pedro and San Luis mesas (basalt-covered tablelands), which are horsts with the San Luis fault zone to the east and the southern Sangre de Cristo fault zone to the west. The map also includes the Sanchez graben (part of the larger Culebra graben), a deep structural basin that lies between the San Luis fault zone (on the west) and the central Sangre de Cristo fault zone (on the east). The oldest rocks exposed in the map area are the Pliocene to upper Oligocene basin-fill sediments of the Santa Fe Group, and Pliocene Servilleta Basalt, a regional series of 3.7?4.8 Ma old flood basalts. Landslide deposits and colluvium that rest on sediments of the Santa Fe Group cover the steep margins of the mesas. Rare exposures of the sediment are comprised of siltstones, sandstones, and minor fluvial conglomerates. Most of the low ground surrounding the mesas and in the graben is covered by surficial deposits of Quaternary age. The alluvial deposits are subdivided into three Pleistocene-age units and three Holocene-age units. The oldest Pleistocene gravel (unit Qao) forms extensive coalesced alluvial fan and piedmont surfaces, the largest of which is known as the Costilla Plain. This surface extends west from San Pedro Mesa to the Rio Grande. The primary geologic hazards in the map area are from earthquakes, landslides, and localized flooding. There are three major fault zones in the area (as discussed above), and they all show evidence for late Pleistocene to possible Holocene movement. The landslides may have seismogenic origins; that is, they may be stimulated by strong ground shaking during large earthquakes. Machette and Thompson based this geologic map entirely on new mapping, whereas Drenth supplied geophysical data and interpretations.

SANS studies of polymers

International Nuclear Information System (INIS)

Wignall, G.D.

1984-10-01

Before small-angle neutron scattering (SANS), chain conformation studies were limited to light and small angle x-ray scattering techniques, usually in dilute solution. SANS from blends of normal and labeled molecules could give direct information on chain conformation in bulk polymers. Water-soluble polymers may be examined in H 2 O/D 2 O mixtures using contrast variation methods to provide further information on polymer structure. This paper reviews some of the information provided by this technique using examples of experiments performed at the National Center for Small-Angle Scattering Research (NCSASR)

Study on effects of development of reactor constant in fast reactor analysis

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

Energy Technology Data Exchange (ETDEWEB)

Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

Butterfly fauna in Mount Gariwang-san, Korea

Directory of Open Access Journals (Sweden)

Cheol Min Lee

2016-06-01

Full Text Available The aim of this study is to elucidate butterfly fauna in Mt. Gariwang-san, Korea. A field survey was conducted from 2010 to 2015 using the line transect method. A literature survey was also conducted. A total of 2,037 butterflies belonging to 105 species were recorded. In the estimation of species richness of butterfly, 116 species were estimated to live in Mt. Gariwang-san. In butterfly fauna in Mt. Gariwang-san, the percentage of northern species was very high and the percentage of grassland species was relatively higher than that of forest edge species and forest interior species. Sixteen red list species were found. In particular, Mimathyma nycteis was only recorded in Mt. Gariwang-san. When comparing the percentage of northern species and southern species including those recorded in previous studies, the percentage of northern species was found to have decreased significantly whereas that of southern species increased. We suggest that the butterfly community, which is distributed at relatively high altitudes on Mt. Gariwang-san, will gradually change in response to climate change.

Guide for monitoring effectiveness of utility Reliability Centered Maintenance (RCM) programs

International Nuclear Information System (INIS)

Midgett, W.D.; Wilson, J.F.; Krochmal, D.F.; Owsenek, L.W.

1991-02-01

Reliability centered maintenance (RCM) programs help utilities optimize preventive maintenance efforts while improving plant safety and economy through increased dependability of plant components. The project team developed this guide and accompanying methodology based on status updates from the Ginna and San Onofre demonstration projects. These updates addressed areas ranging from system selection to the effectiveness of RCM program implementation. In addition, the team incorporated information from a 12-utility survey soliciting opinions on the need for a methodology to monitor RCM cost-effectiveness. An analysis of the 12-utility survey showed that no techniques had been developed to measure RCM program cost-effectiveness. Thus, this guide addresses two key areas: Pros and cons of various monitoring techniques available to assess the overall effectiveness of RCM and a methodology for specifically evaluating the cost-effectiveness of RCM programs. 1 fig

Plant-specific evaluations of Transamerica Delaval diesel engines for nuclear service

International Nuclear Information System (INIS)

Dingee, D.A.; Laity, W.W.; Nesbitt, J.F.

1985-03-01

This paper discusses the approach taken to evlauate the readiness of Transamerica Delaval, Inc. (TDI) diesel generators for nuclear service at five power plants: Catawba, Comanche Peak, Grand Gulf, San Onofre, and Shoreham. TDI engines in these and other nuclear power plants have been the subject of a coordinated effort by 13 nuclear utilities to address reliability and quality issues. The utilities formed the TDI Diesel Generator Owners' Group and prepared a comprehensive plan for requalifying the engines as emergency power sources. Prior to full implementation of the plan by the Owners' Group and final review of the findings by the US Nuclear Regulatory Commission, several member plants became candidates for operating licenses. The TDI engines in those plants, including the five listed above, were evaluated on a case-by-case basis, taking into consideration the factors discussed in this paper. 2 refs

Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

International Nuclear Information System (INIS)

2015-01-01

This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

San Francisco Bay Long Term Management Strategy for Dredging

Science.gov (United States)

The San Francisco Bay Long Term Management Strategy (LTMS) is a cooperative effort to develop a new approach to dredging and dredged material disposal in the San Francisco Bay area. The LTMS serves as the Regional Dredging Team for the San Francisco area.

United States Air Force Personalized Medicine and Advanced Diagnostics Program Panel: Representative Research at the San Antonio Military Medical Center

Science.gov (United States)

2016-05-20

health system. dedicated to excellence in global care PROCESSING OF PROFESSIONAL MEDICAL RESEARCH PUBLICATIONS/PRESENTATIONS INSTRUCTIONS 1. The...present this research at the University of Texas at San Antonio/SAMHS & Universities Research Forum, SURF 2016 in San Antonio, TX, on 20 May 2016. The...at San Antonio/SAMHS & Universities Research Forum, SURF 2016 in San Antonio, TX, on 20 May 2016. 3. LAWS AND REGULATIONS: DoD 5500.07-R, Joint

Detection of tritium in the CO{sub 2} of the reactors G2/G3 using gas chromatography; La detection du tritium par chromatographie gazeuse dans le CO{sub 2} des piles G2/G3

Energy Technology Data Exchange (ETDEWEB)

Guillermin, P; Rossi, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

This gas-phase chromatographic method, based on the principle of the decomposition of a gas mixture into its pure constituents, makes it possible to identify and rapidly measure the tritium present in the heat-carrying fluid of the reactors G2/G3. The sensitivity limit corresponds to 5 x 10{sup -6} {mu}Ci/cm{sup 3} of tritiated gas, whereas the threshold reading of the D.C.C.A. is 10{sup -3} {mu}Ci/cm{sup 3} in the presence of {sup 41}A. This apparatus has interesting applications in the conditions where certain {beta} emitters (products of fission or of activation) interfere with the measurement of the tritium. It can easily be adapted to the detection of tritiated steam on condition that a reducing chemical treatment is applied for the atmospheric humidity. In fact, although this method is not as sensitive for the measurement of tritiated vapour as p-spectrometry in a scintillating medium, it may be set up very easily for measuring the C.M.A of tritium in air and is not affected by the presence of radio-active gases. (authors) [French] Cette methode de chromatographie en phase gazeuse, basee sur le principe de decomposition d'un melange gazeux en ses constituants purs, permet l'identification et la mesure rapide du tritium present dans le fluide caloporteur des piles G2/G3. La limite de sensibilite correspond a 5.10{sup -6} {mu}Ci/cm{sup 3} de gaz tritie, alors que le seuil de lecture du D.C.C.A. s'eleve a 10{sup -3} {mu}Ci/cm{sup 3} en presence de {sup 41}A. Cet appareillage presente un champ d'application interessant dans les domaines ou certains emetteurs {beta} (produits de fission ou d'activation) genent la mesure du tritium. Il peut s'adapter sans difficulte a la detection de la vapeur tritiee moyennant un traitement chimique reducteur de l'humidite atmospherique. En definitive, bien que cette methode ne soit pas aussi sensible pour la determination de la vapeur tritiee que la spectrometrie {beta} en milieu scintillant, elle permet de mesurer la C.M.A de

Direct interaction of the Usher syndrome 1G protein SANS and myomegalin in the retina.

Science.gov (United States)

Overlack, Nora; Kilic, Dilek; Bauss, Katharina; Märker, Tina; Kremer, Hannie; van Wijk, Erwin; Wolfrum, Uwe

2011-10-01

The human Usher syndrome (USH) is the most frequent cause of combined hereditary deaf-blindness. USH is genetically heterogeneous with at least 11 chromosomal loci assigned to 3 clinical types, USH1-3. We have previously demonstrated that all USH1 and 2 proteins in the eye and the inner ear are organized into protein networks by scaffold proteins. This has contributed essentially to our current understanding of the function of USH proteins and explains why defects in proteins of different families cause very similar phenotypes. We have previously shown that the USH1G protein SANS (scaffold protein containing ankyrin repeats and SAM domain) contributes to the periciliary protein network in retinal photoreceptor cells. This study aimed to further elucidate the role of SANS by identifying novel interaction partners. In yeast two-hybrid screens of retinal cDNA libraries we identified 30 novel putative interacting proteins binding to the central domain of SANS (CENT). We confirmed the direct binding of the phosphodiesterase 4D interacting protein (PDE4DIP), a Golgi associated protein synonymously named myomegalin, to the CENT domain of SANS by independent assays. Correlative immunohistochemical and electron microscopic analyses showed a co-localization of SANS and myomegalin in mammalian photoreceptor cells in close association with microtubules. Based on the present results we propose a role of the SANS-myomegalin complex in microtubule-dependent inner segment cargo transport towards the ciliary base of photoreceptor cells. Copyright © 2011 Elsevier B.V. All rights reserved.

Temporal and spatial trends in streamwater nitrate concentrations in the San Bernardino mountains, southern California

Science.gov (United States)

Mark E. Fenn; Mark A. Poth

1999-01-01

We report streamwater nitrate (NO,) concentrations for December 1995 to September 1998 from 19 sampling sites across a N deposition gradient in the San Bernardino Mountains. Streamwater NO3- concentrations in Devil Canyon (DC), a high-pollution area, and in previously reported data from the San Gabriel Mountains 40 km...

San Juan Uchucuanicu: évolution historique

Directory of Open Access Journals (Sweden)

1975-01-01

Full Text Available La communauté de San Juan est reconnue depuis 1939. Une première partie concerne l’organisation de la reducción de San Juan vers le milieu du XVIe siècle. Le poids fiscal s’exerce durement sur le village et la crise est générale dans toute la vallée du Chancay au XVIIe. siècle. La christianisation des habitants est définitive au milieu de ce même siècle. C’est vers la fin du XVIIe siècle et durant tout le XVIIIe que se multiplient les conflits entre San Juan et les villages voisins liés aux terrains de pâture et à la possession de l’eau. La deuxième partie du travail concerne les rapports de la communauté de San Juan avec le Pérou contemporain : contrainte fiscale toujours très lourde durant la fin de l’époque coloniale, exactions des militaires juste avant l’indépendance. La période républicaine voit toujours les conflits avec les villages voisins mais aussi la naissance de familles qui cherchent à retirer le maximum de la communauté. Les terres sont divisées et attribuées : la détérioration de l’organisation communale traditionnelle est manifeste. L4es conflits se multiplient entre petits propriétaires, mais aussi avec les haciendas voisines : c’est l’apparition d’une véritable lutte de classes. La situation actuelle est incertaine, le poids de l’économie marchande se développe avec l’exode des jeunes. Que sera la communauté San Juan à la fin de ce siècle? La comunidad de San Juan está reconocida desde 1939. La primera parte concierne a la organización de la 'reducción' de San Juan hacia mediados del siglo XVI. El peso fiscal se ejerce duramente sobre el pueblo y en el siglo XVII la crisis es general en todo el valle de Chancay. Hacia mediados del mismo siglo la cristianización de los habitantes es definitiva. Es hacia fines del siglo XVII y durante todo el siglo XVIII que se multiplican los conflictos entre San Juan y los pueblos vecinos, los que están relacionados con los terrenos de

Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

Directory of Open Access Journals (Sweden)

Guangcan Zhu

2015-01-01

Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

San Francisco Bay Water Quality Improvement Fund

Science.gov (United States)

EPAs grant program to protect and restore San Francisco Bay. The San Francisco Bay Water Quality Improvement Fund (SFBWQIF) has invested in 58 projects along with 70 partners contributing to restore wetlands, water quality, and reduce polluted runoff.,

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

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

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Development of 40m SANS and Its Utilization Techniques

International Nuclear Information System (INIS)

Choi, Sung Min; Kim, Tae Hwan

2010-06-01

Small angle neutron scattering (SANS) has been a very powerful tool to study nanoscale (1-100 nm) bulk structures in various materials such as polymer, self assembled materials, nano-porous materials, nano-magnetic materials, metal and ceramics. Understanding the importance of the SANS instrument, the 8m SANS instrument was installed at the CN beam port of HANARO in 2001. However, without having a cold neutron source, the beam intensity is fairly low and the Q-range is rather limited due to short instrument length. In July 1, 2003, therefore, the HANARO cold neutron research facility project was launched and a state of the art 40m SANS instrument was selected as top-priority instrument. The development of the 40m SANS instrument was completed as a joint project between Korea Advanced Institute of Science and Technology and the HANARO in 2010. Here, we report the specification of a state of art 40m SANS instrument at HANARO

San Diego's High School Dropout Crisis

Science.gov (United States)

Wilson, James C.

2012-01-01

This article highlights San Diego's dropout problem and how much it's costing the city and the state. Most San Diegans do not realize the enormous impact high school dropouts on their city. The California Dropout Research Project, located at the University of California at Santa Barbara, has estimated the lifetime cost of one class or cohort of…

Coulomb Stress Accumulation along the San Andreas Fault System

Science.gov (United States)

Smith, Bridget; Sandwell, David

2003-01-01

Stress accumulation rates along the primary segments of the San Andreas Fault system are computed using a three-dimensional (3-D) elastic half-space model with realistic fault geometry. The model is developed in the Fourier domain by solving for the response of an elastic half-space due to a point vector body force and analytically integrating the force from a locking depth to infinite depth. This approach is then applied to the San Andreas Fault system using published slip rates along 18 major fault strands of the fault zone. GPS-derived horizontal velocity measurements spanning the entire 1700 x 200 km region are then used to solve for apparent locking depth along each primary fault segment. This simple model fits remarkably well (2.43 mm/yr RMS misfit), although some discrepancies occur in the Eastern California Shear Zone. The model also predicts vertical uplift and subsidence rates that are in agreement with independent geologic and geodetic estimates. In addition, shear and normal stresses along the major fault strands are used to compute Coulomb stress accumulation rate. As a result, we find earthquake recurrence intervals along the San Andreas Fault system to be inversely proportional to Coulomb stress accumulation rate, in agreement with typical coseismic stress drops of 1 - 10 MPa. This 3-D deformation model can ultimately be extended to include both time-dependent forcing and viscoelastic response.

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

Potential field studies of the central San Luis Basin and San Juan Mountains, Colorado and New Mexico, and southern and western Afghanistan

Science.gov (United States)

Drenth, Benjamin John

This dissertation includes three separate chapters, each demonstrating the interpretive utility of potential field (gravity and magnetic) geophysical datasets at various scales and in various geologic environments. The locations of these studies are the central San Luis Basin of Colorado and New Mexico, the San Juan Mountains of southwestern Colorado, and southern and western Afghanistan. The San Luis Basin is the northernmost of the major basins that make up the Rio Grande rift, and interpretation of gravity and aeromagnetic data reveals patterns of rifting, rift-sediment thicknesses, distribution of pre-rift volcanic and sedimentary rocks, and distribution of syn-rift volcanic rocks. Syn-rift Santa Fe Group sediments have a maximum thickness of ˜2 km in the Sanchez graben near the eastern margin of the basin along the central Sangre de Cristo fault zone. Under the Costilla Plains, thickness of these sediments is estimated to reach ˜1.3 km. The Santa Fe Group sediments also reach a thickness of nearly 1 km within the Monte Vista graben near the western basin margin along the San Juan Mountains. A narrow, north-south-trending structural high beneath San Pedro Mesa separates the graben from the structural depression beneath the Costilla Plains. Aeromagnetic anomalies are interpreted to mainly reflect variations of remanent magnetic polarity and burial depth of the 5.3-3.7 Ma Servilleta basalt of the Taos Plateau volcanic field. Magnetic-source depth estimates indicate patterns of subsidence following eruption of the basalt and show that the Sanchez graben has been the site of maximum subsidence. One of the largest and most pronounced gravity lows in North America lies over the rugged San Juan Mountains in southwestern Colorado. A buried, low-density silicic batholith related to an Oligocene volcanic field coincident with the San Juan Mountains has been the accepted interpretation of the source of the gravity low since the 1970s. However, this interpretation was

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

International Nuclear Information System (INIS)

Greene, S.R.; Fisher, S.E.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

Il matrimonio same sex nella Repubblica di San Marino?

Directory of Open Access Journals (Sweden)

Luca Iannaccone

2014-07-01

Full Text Available Contributo sottoposto a valutazioneSOMMARIO: 1. Premessa – 2. La questione del matrimonio tra persone dello stesso sesso in Italia: brevi cenni circa lo stato del dibattito nella giurisprudenza - 3. Il matrimonio same sex nella Repubblica di San Marino?

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

Cenobios leoneses altomedievales ante la europeización: San Pedro y San Pablo de Montes, Santiago y San Martín de Peñalba y San Miguel de Escalada

Directory of Open Access Journals (Sweden)

Martínez Tejera, Artemio Manuel

2002-06-01

Full Text Available The following paper analyses the behaviour of three of the most important monastic communities in the reing of Asturias-Leon for the ninth and then centuries. During this period we witness the implementation of a new ordo, or liturgical ritual that replaces the Hispanic one, strongly established in the Territorium. The liturgical adaptation produces tension and conflicts among the members of different monastic communities, and even between the Episcopate and the monarchy - being King Alfonso VI. In some of the monasteries, the arrival of the new ordo causes the adaptation of the liturgical space, with subsequent changes in liturgical furniture.

El presente estudio pretende analizar el comportamiento de tres de las más importantes comunidades monásticas astur-leonesas de los siglos IX y X (San Pedro y San Pablo de Montes, Santiago y San Martín de Peñalba y San Miguel de Escalada ante la recepción e implantación de aquel nuevo ordo o ritual litúrgico que vino a sustituir al Hispánico, fuertemente asentado en el territorium. Readaptación litúrgica que, con distinta intensidad, producirá tensiones y enfrentamientos entre los miembros de las distintas comunidades monásticas, incluso entre el episcopado y la monarquía (personificada en la figura de Alfonso VI, pero no únicamente. En alguno de estos monasterios la llegada del nuevo ordo supondrá, además, la readaptación de su espacio litúrgico, lo que trajo consigo significativas modificaciones constructivas.

Phosphorylation of the Usher syndrome 1G protein SANS controls Magi2-mediated endocytosis.

Science.gov (United States)

Bauß, Katharina; Knapp, Barbara; Jores, Pia; Roepman, Ronald; Kremer, Hannie; Wijk, Erwin V; Märker, Tina; Wolfrum, Uwe

2014-08-01

The human Usher syndrome (USH) is a complex ciliopathy with at least 12 chromosomal loci assigned to three clinical subtypes, USH1-3. The heterogeneous USH proteins are organized into protein networks. Here, we identified Magi2 (membrane-associated guanylate kinase inverted-2) as a new component of the USH protein interactome, binding to the multifunctional scaffold protein SANS (USH1G). We showed that the SANS-Magi2 complex assembly is regulated by the phosphorylation of an internal PDZ-binding motif in the sterile alpha motif domain of SANS by the protein kinase CK2. We affirmed Magi2's role in receptor-mediated, clathrin-dependent endocytosis and showed that phosphorylated SANS tightly regulates Magi2-mediated endocytosis. Specific depletions by RNAi revealed that SANS and Magi2-mediated endocytosis regulates aspects of ciliogenesis. Furthermore, we demonstrated the localization of the SANS-Magi2 complex in the periciliary membrane complex facing the ciliary pocket of retinal photoreceptor cells in situ. Our data suggest that endocytotic processes may not only contribute to photoreceptor cell homeostasis but also counterbalance the periciliary membrane delivery accompanying the exocytosis processes for the cargo vesicle delivery. In USH1G patients, mutations in SANS eliminate Magi2 binding and thereby deregulate endocytosis, lead to defective ciliary transport modules and ultimately disrupt photoreceptor cell function inducing retinal degeneration. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

Identifying clinically meaningful symptom response cut-off values on the SANS in predominant negative symptoms.

Science.gov (United States)

Levine, Stephen Z; Leucht, Stefan

2013-04-01

The treatment and measurement of negative symptoms are currently at issue in schizophrenia, but the clinical meaning of symptom severity and change is unclear. To offer a clinically meaningful interpretation of severity and change scores on the Scale for the Assessment of Negative Symptoms (SANS). Patients were intention-to-treat participants (n=383) in two double-blind randomized placebo-controlled clinical trials that compared amisulpride with placebo for the treatment of predominant negative symptoms. Equipercentile linking was used to examine extrapolation from (a) CGI-S to SANS severity ratings, and (b) CGI-I to SANS percentage change (n=383). Linking was conducted at baseline, 8-14 days, 28-30 days, and 56-60 days of the trials. Across visits, CGI-S ratings of 'not ill' linked to SANS scores of 0-13, and ranged to 'extreme' ratings that linked to SANS scores of 102-105. The relationship between the CGI-S and the SANS severity scores assumed a linear trend (1=0-13, 2=15-56, 3=37-61, 4=49-66, 5=63-75, 6=79-89, 7=102-105). Similarly the relationship between CGI-I ratings and SANS percentage change followed a linear trend. For instance, CGI-I ratings of 'very much improved' were linked to SANS percent changes of -90 to -67, 'much improved' to -50 to -42, and 'minimally improved' to -21 to -13. The current results uniquely contribute to the debate surrounding negative symptoms by providing clinical meaning to SANS severity and change scores and so offer direction regarding clinically meaningful response cut-off scores to guide treatment targets of predominant negative symptoms. Copyright © 2013 Elsevier B.V. All rights reserved.

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

Archiving and retrieval of experimental data using SAN based centralized storage system for SST-1

Energy Technology Data Exchange (ETDEWEB)

Bhandarkar, Manisha, E-mail: manisha@ipr.res.in; Masand, Harish; Kumar, Aveg; Patel, Kirit; Dhongde, Jasraj; Gulati, Hitesh; Mahajan, Kirti; Chudasama, Hitesh; Pradhan, Subrata

2016-11-15

Highlights: • SAN (Storage Area Network) based centralized data storage system of SST-1 has envisaged to address the need of centrally availability of SST-1 storage system to archive/retrieve experimental data for the authenticated users for 24 × 7. • The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS cluster file system with multipath support. • The adopted SAN based data storage for SST-1 is a modular, robust, and allows future expandability. • Important considerations has been taken like, Handling of varied Data writing speed from different subsystems to central storage, Simultaneous read access of the bulk experimental and as well as essential diagnostic data, The life expectancy of data, How often data will be retrieved and how fast it will be needed, How much historical data should be maintained at storage. - Abstract: SAN (Storage Area Network, a high-speed, block level storage device) based centralized data storage system of SST-1 (Steady State superconducting Tokamak) has envisaged to address the need of availability of SST-1 operation & experimental data centrally for archival as well as retrieval [2]. Considering the initial data volume requirement, ∼10 TB (Terabytes) capacity of SAN based data storage system has configured/installed with optical fiber backbone with compatibility considerations of existing Ethernet network of SST-1. The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS (Global File System) cluster file system with multipath support. Tier-1 is of ∼3 TB (frequent access and low data storage capacity) comprises of Fiber channel (FC) based hard disks for optimum throughput. Tier-2 is of ∼6 TB (less frequent access and high data storage capacity) comprises of SATA based hard disks. Tier-3 will be planned later to store offline historical data. In the SAN configuration two tightly coupled storage servers (with cluster configuration) are

Archiving and retrieval of experimental data using SAN based centralized storage system for SST-1

International Nuclear Information System (INIS)

Bhandarkar, Manisha; Masand, Harish; Kumar, Aveg; Patel, Kirit; Dhongde, Jasraj; Gulati, Hitesh; Mahajan, Kirti; Chudasama, Hitesh; Pradhan, Subrata

2016-01-01

Highlights: • SAN (Storage Area Network) based centralized data storage system of SST-1 has envisaged to address the need of centrally availability of SST-1 storage system to archive/retrieve experimental data for the authenticated users for 24 × 7. • The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS cluster file system with multipath support. • The adopted SAN based data storage for SST-1 is a modular, robust, and allows future expandability. • Important considerations has been taken like, Handling of varied Data writing speed from different subsystems to central storage, Simultaneous read access of the bulk experimental and as well as essential diagnostic data, The life expectancy of data, How often data will be retrieved and how fast it will be needed, How much historical data should be maintained at storage. - Abstract: SAN (Storage Area Network, a high-speed, block level storage device) based centralized data storage system of SST-1 (Steady State superconducting Tokamak) has envisaged to address the need of availability of SST-1 operation & experimental data centrally for archival as well as retrieval [2]. Considering the initial data volume requirement, ∼10 TB (Terabytes) capacity of SAN based data storage system has configured/installed with optical fiber backbone with compatibility considerations of existing Ethernet network of SST-1. The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS (Global File System) cluster file system with multipath support. Tier-1 is of ∼3 TB (frequent access and low data storage capacity) comprises of Fiber channel (FC) based hard disks for optimum throughput. Tier-2 is of ∼6 TB (less frequent access and high data storage capacity) comprises of SATA based hard disks. Tier-3 will be planned later to store offline historical data. In the SAN configuration two tightly coupled storage servers (with cluster configuration) are

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

Stenocercus doellojuradoi (Iguanidae, Tropidurinae): una nueva especie para la provincia de San Juan, Argentina

OpenAIRE

Laspiur, Alejandro; Acosta, Juan Carlos

2006-01-01

República Argentina, Provincia de San Juan, Depto. Valle Fértil, 3 km al norte de la localidad de Las Tumanas sobre la Ruta Provincial 510 (30°52’ S, 67°20’ W). COLECTOR: Alejandro Laspiur. FECHA: 25 /02/ 2006. MATERIAL DE REFERENCIA: Instituto y Museo de Ciencias Naturales, Universidad Nacional de San Juan: IMCNUNSJ 3000. Un ejemplar macho (LHC: 54 mm.).

77 FR 46115 - Notice of Inventory Completion: San Diego Museum of Man, San Diego, CA

Science.gov (United States)

2012-08-02

...The San Diego Museum of Man has completed an inventory of human remains in consultation with the appropriate Indian tribe, and has determined that there is a cultural affiliation between the human remains and a present-day Indian tribe. Representatives of any Indian tribe that believes itself to be culturally affiliated with the human remains may contact the San Diego Museum of Man. Repatriation of the human remains to the Indian tribe stated below may occur if no additional claimants come forward.

Effects of Choto-san and Chotoko on thiopental-induced sleeping time

OpenAIRE

JEENAPONGSA, Rattima; Tohda, Michihisa; Watanabe, Hiroshi

2003-01-01

Choto-san has been used for treatment of centrally regulated disorders such as dementia, hypertension, headache and vertigo. Our laboratory showed that Choto-san improved learning memory in ischemic mice. It is noticeable that Choto-san treated animals and animals that underwent conducting occlusion of common carotid arteries (2VO) operation slept longer than the normal animals. Therefore, this study aimed to clarify the effects of Choto-san and its related component; Chotoko and Choto-san wi...

Impact of a Large San Andreas Fault Earthquake on Tall Buildings in Southern California

Science.gov (United States)

Krishnan, S.; Ji, C.; Komatitsch, D.; Tromp, J.

2004-12-01

In 1857, an earthquake of magnitude 7.9 occurred on the San Andreas fault, starting at Parkfield and rupturing in a southeasterly direction for more than 300~km. Such a unilateral rupture produces significant directivity toward the San Fernando and Los Angeles basins. The strong shaking in the basins due to this earthquake would have had a significant long-period content (2--8~s). If such motions were to happen today, they could have a serious impact on tall buildings in Southern California. In order to study the effects of large San Andreas fault earthquakes on tall buildings in Southern California, we use the finite source of the magnitude 7.9 2001 Denali fault earthquake in Alaska and map it onto the San Andreas fault with the rupture originating at Parkfield and proceeding southward over a distance of 290~km. Using the SPECFEM3D spectral element seismic wave propagation code, we simulate a Denali-like earthquake on the San Andreas fault and compute ground motions at sites located on a grid with a 2.5--5.0~km spacing in the greater Southern California region. We subsequently analyze 3D structural models of an existing tall steel building designed in 1984 as well as one designed according to the current building code (Uniform Building Code, 1997) subjected to the computed ground motion. We use a sophisticated nonlinear building analysis program, FRAME3D, that has the ability to simulate damage in buildings due to three-component ground motion. We summarize the performance of these structural models on contour maps of carefully selected structural performance indices. This study could benefit the city in laying out emergency response strategies in the event of an earthquake on the San Andreas fault, in undertaking appropriate retrofit measures for tall buildings, and in formulating zoning regulations for new construction. In addition, the study would provide risk data associated with existing and new construction to insurance companies, real estate developers, and

33 CFR 110.120 - San Luis Obispo Bay, Calif.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false San Luis Obispo Bay, Calif. 110... ANCHORAGES ANCHORAGE REGULATIONS Special Anchorage Areas § 110.120 San Luis Obispo Bay, Calif. (a) Area A-1. Area A-1 is the water area bounded by the San Luis Obispo County wharf, the shoreline, a line drawn...

Species Observations (poly) - San Diego County [ds648

Data.gov (United States)

California Natural Resource Agency — Created in 2009, the SanBIOS database serves as a single repository of species observations collected by various departments within the County of San Diego's Land...

Mammal Track Counts - San Diego County [ds442

Data.gov (United States)

California Natural Resource Agency — The San Diego Tracking Team (SDTT) is a non-profit organization dedicated to promoting the preservation of wildlife habitat in San Diego County through citizen-based...

Species Observations (poly) - San Diego County [ds648

Data.gov (United States)

California Department of Resources — Created in 2009, the SanBIOS database serves as a single repository of species observations collected by various departments within the County of San Diego's Land...

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

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

Biological and associated water-quality data for lower Olmos Creek and upper San Antonio River, San Antonio, Texas, March-October 1990

Science.gov (United States)

Taylor, R. Lynn

1995-01-01

Biological and associated water-quality data were collected from lower Olmos Creek and upper San Antonio River in San Antonio, Texas, during March-October 1990, the second year of a multiyear data-collection program. The data will be used to document water-quality conditions prior to implementation of a proposal to reuse treated wastewater to irrigate city properties in Olmos Basin and Brackenridge Parks and to augment flows in the Olmos Creek/San Antonio River system.

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

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

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

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

Reactor safety study applied to the Forsmark 3 Power Plant

International Nuclear Information System (INIS)

Ericsson, G.; Tiren, L.I.

1978-01-01

A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

Remembering San Diego

International Nuclear Information System (INIS)

Chuyanov, V.

1999-01-01

After 6 years of existence the ITER EDA project in San Diego, USA, was terminated by desition of the US Congress. This article describes how nice it was for everybody as long as it lasted and how sad it is now

Tratamiento biológico del lixiviado generado en el relleno sanitario "El Guayabal" de la ciudad San José de Cúcuta

OpenAIRE

Alexander Álvarez Contreras; John Hermógenes Suárez Gelvez

2006-01-01

En este trabajo se realizó un diagnóstico de calidad y cantidad del lixiviado generado en el relleno sanitario El Guayabal de la ciudad San José de Cúcuta, y se evaluaron dos sistemas de tratamiento biológico a escala laboratorio para este lixiviado. El lixiviado en el momento de la experiencia presentaba un rango de DQO de 7.650 a 28.250 mg/L. Los sistemas de tratamiento ensayados fueron: un reactor anaerobio del tipo UASB y un sistema de Biodiscos. La carga máxima asimilada p...

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

Update: San Andreas Fault experiment

Science.gov (United States)

Christodoulidis, D. C.; Smith, D. E.

1984-01-01

Satellite laser ranging techniques are used to monitor the broad motion of the tectonic plates comprising the San Andreas Fault System. The San Andreas Fault Experiment, (SAFE), has progressed through the upgrades made to laser system hardware and an improvement in the modeling capabilities of the spaceborne laser targets. Of special note is the launch of the Laser Geodynamic Satellite, LAGEOS spacecraft, NASA's only completely dedicated laser satellite in 1976. The results of plate motion projected into this 896 km measured line over the past eleven years are summarized and intercompared.

Vabariigi aastapäev San Franciscos / Heino Valvur ; foto: Heino Valvur

Index Scriptorium Estoniae

Valvur, Heino

2006-01-01

veebruarikuu möödus San Franciscos Eesti Vabariigi 88. aastapäeva pühitsedes: traditsiooniliselt tähistas aastapäeva San Francisco Seenioride Klubi koosviibimisega, E.E.L.K. San Francisco koguduses peeti jumalateenistus ja koosviibimine, kus noored esitasid rahvalaule, San Francisco Eesti Selts tähistas aastapäeva 25. veebruaril aktuse ja koosviibimisega

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)

Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Arigane, Kenji

2005-03-01

Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

76 FR 70480 - Otay River Estuary Restoration Project, South San Diego Bay Unit of the San Diego Bay National...

Science.gov (United States)

2011-11-14

... River Estuary Restoration Project, South San Diego Bay Unit of the San Diego Bay National Wildlife...), intend to prepare an environmental impact statement (EIS) for the proposed Otay River Estuary Restoration... any one of the following methods. Email: [email protected] . Please include ``Otay Estuary NOI'' in the...

Stenocercus doellojuradoi (Iguanidae, Tropidurinae: una nueva especie para la provincia de San Juan, Argentina

Directory of Open Access Journals (Sweden)

Laspiur, Alejandro

2006-05-01

Full Text Available República Argentina, Provincia de San Juan, Depto. Valle Fértil, 3 km al norte de la localidad de Las Tumanas sobre la Ruta Provincial 510 (30°52’ S, 67°20’ W. COLECTOR: Alejandro Laspiur. FECHA: 25 /02/ 2006. MATERIAL DE REFERENCIA: Instituto y Museo de Ciencias Naturales, Universidad Nacional de San Juan: IMCNUNSJ 3000. Un ejemplar macho (LHC: 54 mm..

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

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

Analytical models for lower and upper bounds of the condensation-induced water hammer in long horizontal pipes

International Nuclear Information System (INIS)

Chun, Moon Hyun; Park, Joo Wan; Nam, Ho Yun

1992-01-01

Improved analytical models have been proposed that can predict the lower and upper limits of the water hammer region for given flow conditions by incorporation of recent advances made in the understanding of phenomena associated with the condensation-induced water hammer into existing methods. Present models are applicable for steam-water counterflow in a long horizontal pipe geometry. Both lower and upper bounds of the water hammer region are expressed in terms of the 'critical inlet water flow rate' as a function of axial position. Water hammer region boundaries predicted by present and typical existing models are compared for particular flow conditions of the water hammer event occurred at San Onofre Unit 1 to assess the applicability of the models examined. The result shows that present models for lower and upper bounds of the water hammer region compare favorably with the best performing existing models

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Lessons learned from the SONGS Unit 1 water hammer event

International Nuclear Information System (INIS)

Chiu, C.

1987-01-01

On November 21, 1985, a water hammer event occurred in horizontal feedwater line B at San Onofre Nuclear Generation Site (SONGS) Unit 1. The SONGS Unit 1 is a three-loop pressurized water reactor designed by Westinghouse Electric Corp. The event was initiated by a differential current trip on the bus of auxiliary transformer C. The root cause of the event was a simultaneous failure of five check valves in the feedwater system. Two of them are located downstream of the feedwater pump, and three of them are located further downstream and on the lines to the steam generators. The failure mechanism was determined to be flow-induced vibration, which caused repeated impact between the disk stud and the disk stop. The water hammer occurred in feedwater line B during the refilling of feedwater lines A, B, and C with auxiliary feedwater. The thermal-hydraulic process to initiate the water hammer and the reason that the water hammer only happened in line B have been fully investigated and explained. A root cause analysis after the event was prompted to answer the following two questions: (1) why did these five check valves fail at that time and not in the preceding 15 yr? (2) why did only these five check valves fail? The scope of the root cause analysis involves an investigation of the valve vibration characteristics, plant operation history, and the maintenance history of the valves. The paper answers these two questions, after a brief study of the vibration characteristics of a check valve

Revision of fast reactor group constant set JFS-3-J2

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

Backwater Flooding in San Marcos, TX from the Blanco River

Science.gov (United States)

Earl, Richard; Gaenzle, Kyle G.; Hollier, Andi B.

2016-01-01

Large sections of San Marcos, TX were flooded in Oct. 1998, May 2015, and Oct. 2015. Much of the flooding in Oct. 1998 and Oct. 2015 was produced by overbank flooding of San Marcos River and its tributaries by spills from upstream dams. The May 2015 flooding was almost entirely produced by backwater flooding from the Blanco River whose confluence is approximately 2.2 miles southeast of downtown. We use the stage height of the Blanco River to generate maps of the areas of San Marcos that are lower than the flood peaks and compare those results with data for the observed extent of flooding in San Marcos. Our preliminary results suggest that the flooding occurred at locations more than 20 feet lower than the maximum stage height of the Blanco River at San Marcos gage (08171350). This suggest that the datum for either gage 08171350 or 08170500 (San Marcos River at San Marcos) or both are incorrect. There are plans for the U.S. Army Corps of Engineers to construct a Blanco River bypass that will divert Blanco River floodwaters approximately 2 miles farther downstream, but the $60 million price makes its implementation problematic.

33 CFR 165.1187 - Security Zones; Golden Gate Bridge and the San Francisco-Oakland Bay Bridge, San Francisco Bay...

Science.gov (United States)

2010-07-01

... Limited Access Areas Eleventh Coast Guard District § 165.1187 Security Zones; Golden Gate Bridge and the... Golden Gate Bridge and the San Francisco-Oakland Bay Bridge, in San Francisco Bay, California. (b... siren, radio, flashing light, or other means, the operator of a vessel shall proceed as directed. [COTP...

Simulation of SANS signal due to radiation damage in Fe

International Nuclear Information System (INIS)

Yu, G.; Schaublin, R.; Spatig, P.; Fikar, J.; Baluc, N.

2007-01-01

Full text of publication follows: A wide number of irradiation-induced defects in Fe-base materials (e.g. RAFM steels) have sizes in the range about 0.5 to 1 nm, which are expected to contribute to the irradiation-induced hardening and/or embrittlement phenomena. These defects are at the limit in spatial resolution of transmission electron microscopy (TEM), but they can be investigated using the small angle neutron scattering (SANS) technique, at least in terms of number density and size distribution. Determination of the type of defects (small dislocation loops, interstitials or vacancy clusters, precipitates and cavities, like voids or helium bubbles) is not straightforward. In order to analyze the type of nanometer-sized irradiation-induced defects in Fe-base materials Molecular Dynamics (MD) simulations of various distributions of different types irradiation-induced defects have been performed. The defects investigated consisted in dislocation loops with sizes of 0.5, 1.0 and 2.0 nm and 1/2 a 0 , 1/2 a 0 and a 0 Burgers vectors, cavities, like voids and helium bubbles, with sizes of 0.5, 1.0 and 2.0 nm, MD simulations of atomic displacement cascades were performed using MD samples with a size of 18 x 18 x 18 nm 3 at 10, 300 and 523 K, for PKA energies of 1, 3, 7 and 10 keV. Simulation of the corresponding nuclear SANS signal was performed using the Electron Microscopy Software (EMS) code that was originally designed to simulate TEM images and diffraction patterns and that was modified to simulate the SANS signal. Results of such simulations in pure Fe have been compared to experimental SANS measurements and TEM observations of irradiation-induced defects in Fe-base materials. (authors)

Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

International Nuclear Information System (INIS)

Uegata, T.; Saji, E.; Tanaka, H.

1993-01-01

Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

Application of an unstructured 3D finite volume numerical model to flows and salinity dynamics in the San Francisco Bay-Delta

Science.gov (United States)

Martyr-Koller, R.C.; Kernkamp, H.W.J.; Van Dam, Anne A.; Mick van der Wegen,; Lucas, Lisa; Knowles, N.; Jaffe, B.; Fregoso, T.A.

2017-01-01

A linked modeling approach has been undertaken to understand the impacts of climate and infrastructure on aquatic ecology and water quality in the San Francisco Bay-Delta region. The Delft3D Flexible Mesh modeling suite is used in this effort for its 3D hydrodynamics, salinity, temperature and sediment dynamics, phytoplankton and water-quality coupling infrastructure, and linkage to a habitat suitability model. The hydrodynamic model component of the suite is D-Flow FM, a new 3D unstructured finite-volume model based on the Delft3D model. In this paper, D-Flow FM is applied to the San Francisco Bay-Delta to investigate tidal, seasonal and annual dynamics of water levels, river flows and salinity under historical environmental and infrastructural conditions. The model is driven by historical winds, tides, ocean salinity, and river flows, and includes federal, state, and local freshwater withdrawals, and regional gate and barrier operations. The model is calibrated over a 9-month period, and subsequently validated for water levels, flows, and 3D salinity dynamics over a 2 year period.Model performance was quantified using several model assessment metrics and visualized through target diagrams. These metrics indicate that the model accurately estimated water levels, flows, and salinity over wide-ranging tidal and fluvial conditions, and the model can be used to investigate detailed circulation and salinity patterns throughout the Bay-Delta. The hydrodynamics produced through this effort will be used to drive affiliated sediment, phytoplankton, and contaminant hindcast efforts and habitat suitability assessments for fish and bivalves. The modeling framework applied here will serve as a baseline to ultimately shed light on potential ecosystem change over the current century.

Application of an unstructured 3D finite volume numerical model to flows and salinity dynamics in the San Francisco Bay-Delta

Science.gov (United States)

Martyr-Koller, R. C.; Kernkamp, H. W. J.; van Dam, A.; van der Wegen, M.; Lucas, L. V.; Knowles, N.; Jaffe, B.; Fregoso, T. A.

2017-06-01

A linked modeling approach has been undertaken to understand the impacts of climate and infrastructure on aquatic ecology and water quality in the San Francisco Bay-Delta region. The Delft3D Flexible Mesh modeling suite is used in this effort for its 3D hydrodynamics, salinity, temperature and sediment dynamics, phytoplankton and water-quality coupling infrastructure, and linkage to a habitat suitability model. The hydrodynamic model component of the suite is D-Flow FM, a new 3D unstructured finite-volume model based on the Delft3D model. In this paper, D-Flow FM is applied to the San Francisco Bay-Delta to investigate tidal, seasonal and annual dynamics of water levels, river flows and salinity under historical environmental and infrastructural conditions. The model is driven by historical winds, tides, ocean salinity, and river flows, and includes federal, state, and local freshwater withdrawals, and regional gate and barrier operations. The model is calibrated over a 9-month period, and subsequently validated for water levels, flows, and 3D salinity dynamics over a 2 year period. Model performance was quantified using several model assessment metrics and visualized through target diagrams. These metrics indicate that the model accurately estimated water levels, flows, and salinity over wide-ranging tidal and fluvial conditions, and the model can be used to investigate detailed circulation and salinity patterns throughout the Bay-Delta. The hydrodynamics produced through this effort will be used to drive affiliated sediment, phytoplankton, and contaminant hindcast efforts and habitat suitability assessments for fish and bivalves. The modeling framework applied here will serve as a baseline to ultimately shed light on potential ecosystem change over the current century.

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

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

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

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

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

California State Waters Map Series: offshore of San Gregorio, California

Science.gov (United States)

Cochrane, Guy R.; Dartnell, Peter; Greene, H. Gary; Watt, Janet T.; Golden, Nadine E.; Endris, Charles A.; Phillips, Eleyne L.; Hartwell, Stephen R.; Johnson, Samuel Y.; Kvitek, Rikk G.; Erdey, Mercedes D.; Bretz, Carrie K.; Manson, Michael W.; Sliter, Ray W.; Ross, Stephanie L.; Dieter, Bryan E.; Chin, John L.; Cochran, Susan A.; Cochrane, Guy R.; Cochran, Susan A.

2014-01-01

In 2007, the California Ocean Protection Council initiated the California Seafloor Mapping Program (CSMP), designed to create a comprehensive seafloor map of high-resolution bathymetry, marine benthic habitats, and geology within the 3-nautical-mile limit of California's State Waters. The CSMP approach is to create highly detailed seafloor maps through collection, integration, interpretation, and visualization of swath sonar data, acoustic backscatter, seafloor video, seafloor photography, high-resolution seismic-reflection profiles, and bottom-sediment sampling data. The map products display seafloor morphology and character, identify potential marine benthic habitats, and illustrate both the surficial seafloor geology and shallow (to about 100 m) subsurface geology. The Offshore of San Gregorio map area is located in northern California, on the Pacific coast of the San Francisco Peninsula about 50 kilometers south of the Golden Gate. The map area lies offshore of the Santa Cruz Mountains, part of the northwest-trending Coast Ranges that run roughly parallel to the San Andreas Fault Zone. The Santa Cruz Mountains lie between the San Andreas Fault Zone and the San Gregorio Fault system. The nearest significant onshore cultural centers in the map area are San Gregorio and Pescadero, both unincorporated communities with populations well under 1,000. Both communities are situated inland of state beaches that share their names. No harbor facilities are within the Offshore of San Gregorio map area. The hilly coastal area is virtually undeveloped grazing land for sheep and cattle. The coastal geomorphology is controlled by late Pleistocene and Holocene slip in the San Gregorio Fault system. A westward bend in the San Andreas Fault Zone, southeast of the map area, coupled with right-lateral movement along the San Gregorio Fault system have caused regional folding and uplift. The coastal area consists of high coastal bluffs and vertical sea cliffs. Coastal promontories in

Fitting the datum of SANS with Pxy program

International Nuclear Information System (INIS)

Sun, Liangwei; Peng, Mei; Chen, Liang

2009-04-01

The thesis introduces the basic theory of Small-Angle neutron scattering, enumerates several approximate law. It simply describes the components of Small-Angle neutron spectrometer (SANS) and the parameters of SANS of Budapest Neutron Center (BNC) in Hungary. During the period of studying at Budapest Neutron Center in Hungary, the experiments of wavelength calibration was carried out with SIBE and the SANS experiments of sample Micelles. The experiments are briefly introduced. Pxy program is used to fit these datum, and the results of wavelength and sizes of sample Micelles are presented. (authors)

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Energy Technology Data Exchange (ETDEWEB)

C. B. Davis

2006-07-01

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

City of San Francisco, California street tree resource analysis

Science.gov (United States)

E.G. McPherson; J.R. Simpson; P.J. Peper; Q. Xiao

2004-01-01

Street trees in San Francisco are comprised of two distinct populations, those managed by the city’s Department of Public Works (DPW) and those managed by private property owners with or without the help of San Francisco’s urban forestry nonprofit, Friends of the Urban Forest (FUF). These two entities believe that the public’s investment in stewardship of San Francisco...

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Radiation protection at the RA Reactor in 1993, RA research reactor, Part

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

1993-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

2004-01-01

The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

Trouble Brewing in San Francisco. Policy Brief

Science.gov (United States)

Buck, Stuart

2010-01-01

The city of San Francisco will face enormous budgetary pressures from the growing deficits in public pensions, both at a state and local level. In this policy brief, the author estimates that San Francisco faces an aggregate $22.4 billion liability for pensions and retiree health benefits that are underfunded--including $14.1 billion for the city…

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

International Nuclear Information System (INIS)

Hernandez, N.; Alonso, G.; Valle, E. del

2004-01-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

Experimental investigation of high He/dpa microstructural effects in neutron irradiated B-alloyed Eurofer97 steel by means of small angle neutron scattering (SANS and electron microscopy

Directory of Open Access Journals (Sweden)

R. Coppola

2016-12-01

Full Text Available High He/dpa microstructural effects have been investigated, by means of small-angle neutron scattering (SANS and transmission electron microscopy (TEM, in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2V, 1.08W wt%, B contents variable between 10 and 1000ppm, neutron irradiated at the High Flux Reactor of the JRC-Petten at temperatures between 250 °C and 450 °C, up do a dose level of 16 dpa. Under these irradiation parameters, B activation is expected to produce corresponding helium contents variable between 80 and 5600appm, with helium bubble distributions relevant for the technological applications. The SANS measurements were carried out under magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. Increasing the estimated helium content from 400 to 5600appm, the analysis of the SANS cross-sections yields an increase in the volume fraction, attributed to helium bubbles, of almost one order of magnitude (from 0.007 to 0.038; furthermore, the difference between nuclear and magnetic SANS components is strongly reduced. These results are discussed in correlation with TEM observations of the same samples and are tentatively attributed to the effect of drastic microstructural changes in Eurofer97-1 for high He/dpa ratio values, possibly relating to the dissolution of large B-carbides due to transmutation reactions.

Description of the Triton reactor; Pile Triton, rapport descriptif

Energy Technology Data Exchange (ETDEWEB)

NONE

1967-09-01

The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965. [French] Le reacteur Triton est un reacteur piscine, a uranium enrichi. Il est entre en fonctionnement en 1959, apres une divergence effectuee le 30 juin de cette meme annee. Destine a des etudes de protection contre les rayonnements, son coeur pouvait se deplacer dans le sens longitudinal. La piscine peut etre separee en deux compartiments inegaux par un batardeau. Le coeur triton est place dans le petit compartiment, le coeur Nereide dans le grand compartiment. Un troisieme compartiment sans eau, appele Naiade II, est separe par une paroi en beton dans laquelle est amenagee une fenetre obturee par une plaque d'aluminium (2,50 m x 2,70 m). La fosse Naiade II sert a des experiences de protection utilisant le coeur nereide. Apres une refonte complete, la puissance du reacteur triton qui etait passee progressivement de 1,2 MW a 2 MW, puis 3 MW, a atteint en aout 1965 6, 5 MW. Le

Short-period strain (0.1-105 s): Near-source strain field for an earthquake (M L 3.2) near San Juan Bautista, California

Science.gov (United States)

Johnston, M. J. S.; Borcherdt, R. D.; Linde, A. T.

1986-10-01

Measurements of dilational earth strain in the frequency band 25-10-5 Hz have been made on a deep borehole strainmeter installed near the San Andreas fault. These data are used to determine seismic radiation fields during nuclear explosions, teleseisms, local earthquakes, and ground noise during seismically quiet times. Strains of less than 10-10 on these instruments can be clearly resolved at short periods (< 10 s) and are recorded with wide dynamic range digital recorders. This permits measurement of the static and dynamic strain variations in the near field of local earthquakes. Noise spectra for earth strain referenced to 1 (strain)2/Hz show that strain resolution decreases at about 10 dB per decade of frequency from -150 dB at 10-4 Hz to -223 dB at 10 Hz. Exact expressions are derived to relate the volumetric strain and displacement field for a homogeneous P wave in a general viscoelastic solid as observed on colocated dilatometers and seismometers. A rare near-field recording of strain and seismic velocity was obtained on May 26, 1984, from an earthquake (ML 3.2) at a hypocentral distance of 3.2 km near the San Andreas fault at San Juan Bautista, California. While the data indicate no precursory strain release at the 5 × 10-11 strain level, a coseismic strain release of 1.86 nanostrain was observed. This change in strain is consistent with that calculated from a simple dislocation model of the event. Ground displacement spectra, determined from the downhole strain data and instrument-corrected surface seismic data, suggest that source parameters estimated from surface recordings may be contaminated by amplification effects in near-surface low-velocity materials.

Survey of group data libraries for use of the DYN3D program for WWER type reactors

International Nuclear Information System (INIS)

Mittag, S.

1994-06-01

So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

77 FR 66499 - Environmental Impact Statement: San Bernardino and Los Angeles Counties, CA

Science.gov (United States)

2012-11-05

... San Bernardino, 285 East Hospitality Lane, San Bernardino, California 92408 (2) Sheraton Ontario..., November 13, 2012 from 5-7 p.m. at the Hilton San Bernardino, 285 East Hospitality Lane, San Bernardino...

33 CFR 110.74c - Bahia de San Juan, PR.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Bahia de San Juan, PR. 110.74c Section 110.74c Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY ANCHORAGES ANCHORAGE REGULATIONS Special Anchorage Areas § 110.74c Bahia de San Juan, PR. The waters of San Antonio...