Science.gov

Sample records for russian reactor pressure

  1. Launch of Russian reactor postponed

    SciTech Connect

    Not Available

    1993-02-05

    Astronomers and weapons scientists seemed heated on a collision course a few months ago over the military's plans to send a Russian nuclear reactor into space. But an agreement reached in late January has prevented a pile-up, at least for 6 months. The astronomers, led by Donald Lamb of the University of Chicago, were objecting to plans by the Strategic Defense Initiative Office (SDIO) to launch Topaz 2, an experimental Russian nuclear reactor, arguing that rogue particles from it might ruin sensitive gamma ray experiments. The reactor is designed to propel itself in space with a jet of xenon ions. One worry was that leaking gamma rays and positrons, which can travel in the earth's magnetic field and pop up in the darndest places, might cause false signals in gamma ray monitors (Science, 18 December 1992, p. 1878). The worry has abated now that SDI officials will postpone choosing a rocket and mission altitutde for Topaz 2 for 6 months, while experts study how its emissions at various altitudes might affect instruments aboard the Gamma Ray Observatory and other satellites. In effect, the SDIO has agreed to an environmental impact study for space, following an unusual meeting organized by former Russian space official Roald Sagdeev at the University of Maryland on 19 January. There the Russian designers of Topaz 2, its new owners at the SDIO, and critics in the astronomy community achieved common ground: that more study was needed.

  2. Russian RBMK reactor design information

    SciTech Connect

    Not Available

    1993-11-01

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

  3. Evaluating Russian space nuclear reactor technology for United States applications

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-08-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

  4. Russian-American venture designs new reactor

    SciTech Connect

    Newman, P.

    1994-01-03

    Russian and American nuclear energy experts have completed a joint design study of a small, low-cost and demonstrably accident-proof reactor that they say could revolutionize the way conventional reactors are designed, marketed and operated. The joint design is helium-cooled and graphite-moderated and has a power density of 3 MWt/cubic meter, which is significantly less than the standard American reactor. A prototype of this design should be operating in Chelyabinsk by June 1996.

  5. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  6. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  7. Converting Russian plutonium-production reactors to civilian use

    SciTech Connect

    Dmitriev, A.M.

    1994-12-01

    The first Soviet reactor designed to produce weapon-grade plutonium began operation in Chelyabinsk in June of 1948. In 1958, a graphite-moderated production reactor was built to operate in dual-purpose mode, producing both weapons-grade plutonium and heat and electricity for local residents. In 1963, the first and only US dual-purpose reactor was built in Hanford, Washington. The reactor was shut down for upgrade in January 1987 following the Chernobyl accident and its operation has never been resumed. There are major differences between US and Russian dual-purpose reactors. This article outlines those differences and the problems involved in the conversion of the Russian reactors to civilian use.

  8. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  9. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, R.C.; Upton, H.A.

    1994-10-04

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  10. The Effect of Russian Versus Hebrew Instructions on the Reaction to Social Pressure of Russian-Born Israeli Children

    ERIC Educational Resources Information Center

    Kav-Venaki, Sophie; And Others

    1976-01-01

    An experiment on the role of the language of instruction in mediating responses to social pressure was conducted with a sample of 41 Russian-born adolescents who had recently immigrated to Israel. (Editor)

  11. Neutron flux spectra and radiation damage parameters for the Russian Bor-60 and SM-2 reactors

    SciTech Connect

    Karasiov, A.V.; Greenwood, L.R.

    1995-04-01

    The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

  12. Windowless High-Pressure Solar Reactor

    NASA Technical Reports Server (NTRS)

    Ramohalli, K. N. R.

    1985-01-01

    Obscuration by reaction products eliminated. Chemical reactor heated by Sunlight employs rocket technology to maintain internal pressure. Instead of keeping chamber tightly closed, pressure maintained by momentum balance between incoming and outgoing materials. Windowless solar reactor admits concentrated Sunlight through exhaust aperture. Pressure in reactor maintained dynamically.

  13. A comparative risk assessment for the Russian V213 power reactor

    SciTech Connect

    Marshall, T.D.; Hockenbury, R.W.; Honey, J.A.; Cadwallader, L.C.

    1996-04-01

    Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. US-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here are: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, and small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability data from US PWRs and (b) assuming common component reliability data. US PWR component reliability data are used based of the unavailability of such data for the V213 at the time of the analyses. While the use of US PWR data in this risk assessment of the V213 does strongly infer V213 comparability to US plants, the risk assessment using common component reliability does not have such a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor`s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to US PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures.

  14. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect

    Reynolds, E.; Schaefer, E.; Polansky, G.; Lacy, J.; Bocharov, A.

    1993-09-30

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  15. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, James K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  16. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  17. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, J.K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  18. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  19. Reactor pressure vessel. Status report

    SciTech Connect

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

  20. Pressurized water reactor flow skirt apparatus

    DOEpatents

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  1. Reactor pressure vessel with forged nozzles

    DOEpatents

    Desai, Dilip R.

    1993-01-01

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  2. In-core pressure tube rupture at N Reactor

    SciTech Connect

    McLaren, J.M.

    1989-07-01

    A double-ended guillotine break in a pressure tube while the reactor was at full power was analyzed. The analysis showed that the reactor scrams on high pressure tube flow. The released coolant escapes into the reactor graphite moderator where the reactor gas system vents direct the flow to the confinement. There was no fuel damage. 10 refs., 31 figs.

  3. Tritium issues in commercial pressurized water reactors

    SciTech Connect

    Jones, G.

    2008-07-15

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  4. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    SciTech Connect

    Chebeskov, A.; Kalashnikov, A.; Bevard, B.; Moses, D.; Pavlovichev, A.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  5. 98. ARAIII. ML1 reactor pressure vessel is lowered into reactor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    98. ARA-III. ML-1 reactor pressure vessel is lowered into reactor pit by hoist. July 13, 1963. Ineel photo no. 63-4049. Photographer: Lowin. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  6. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  7. Structural integrity of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  8. PKL reactor tank bottom pressures in accident scenarios

    SciTech Connect

    Tudor, A.A.

    1987-03-10

    Nuclear Engineering Division requested estimates of the maximum PKL reactor tank pressures associated with postulated reactor accidents. Tank bottom pressures calculated in establishing confinement protection limits (CPL) in Mark 16B-31 and Mark 22 reactor charges are given in this document.

  9. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    SciTech Connect

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky; Miroslav Picek; Alexey Smirnov; Sergey Komarov; Edward Bradley; Alexander Dudchenko; Konstantin Golubkin

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.

  10. The current state of the Russian reduced enrichment research reactors program

    SciTech Connect

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  11. Reactor pressure vessel stud management automation strategies

    SciTech Connect

    Biach, W.L.; Hill, R.; Hung, K. )

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities.

  12. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect

    Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

    2003-02-26

    Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

  13. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  14. Reactor pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  15. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  16. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  17. Pressurized reactor system and a method of operating the same

    DOEpatents

    Isaksson, Juhani M.

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  18. Midland reactor pressure vessel flaw distribution

    SciTech Connect

    Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center`s (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions.

  19. Neutron shielding panels for reactor pressure vessels

    SciTech Connect

    Singleton, Norman R.

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  20. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    SciTech Connect

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

    1997-02-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  1. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  2. State space modeling of reactor core in a pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  3. Light Water Reactor-Pressure Vessel Surveillance project computer system

    SciTech Connect

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes.

  4. IRRADIATION CREEP AND SWELLING OF RUSSIAN FERRITIC-MARTENSITIC STEELS IRRADIATED TO VERY HIGH EXPOSURES IN THE BN-350 FAST REACTOR AT 305-335 DEGREES C

    SciTech Connect

    Konobeev, Yu V.; Dvoraishin, A. M.; Porollo, S. I.; Shulepin, S. V.; Budylkin, N. I.; Mironova, E. G.; Garner, Francis A.; Toloczko, Mychailo B.

    2003-09-03

    Russian ferritic martensitic (F(slash)M) steels EP(dash)450, EP(dash)852 and EP(dash)823 were irradiated in the BN(dash)350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb(dash)Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP(dash)450 and EP(dash) 823 at temperatures between 390 and 520 degrees C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP(dash)450 and EP(dash)852 at temperatures between 305 and 335 degrees C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation related densification. These irradiation creep studies confirm that the creep compliance of F(slash)M steels is about one half that of austenitic steels.

  5. Pressurized water nuclear reactor system with hot leg vortex mitigator

    SciTech Connect

    Lau, L.K.S.

    1990-09-18

    This patent describes an improvement in a pressurized water nuclear reactor system having a reactor pressure vessel, at least one steam generator, a hot leg conduit for charging of hot fluid from the reactor pressure vessel to the steam generator, and at least one cold leg conduit for return of cool fluid from the steam generator back to the reactor pressure vessel. The improvement comprises a residual heat removal device wherein: the hot leg has an inside diameter D{sub 1}; a first section of residual heat removal conduit is provided, having an inside diameter D{sub 2}, a first end for receipt of fluid from the hot leg, and a second end; a second section of residual heat removal conduit is provided connected to the reactor pressure vessel; a pump interconnects the second end of the first section of residual heat removal conduit with the second section of residual heat removal conduit; and a step nozzle of an inside diameter D{sub 3} and a length L interconnects the hot leg to the first end of the first section of residual heat removal conduit, with D{sub 3}/D{sub 1} {ge} 0.55, with D{sub 3}/D{sub 2}1.9 and L/D{sub 3} {ge} 1.44.

  6. SCW Pressure-Channel Nuclear Reactor Some Design Features

    NASA Astrophysics Data System (ADS)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  7. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    SciTech Connect

    Wang, Jy-An John

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  8. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  9. LPT. EBOR reactor vessel in TAN 646. Pressure vessel head ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    LPT. EBOR reactor vessel in TAN 646. Pressure vessel head being installed in vault. Refueling port extension (right) and control rod nozzles (center). Camera facing northwest. Photographer: Comiskey. Date: January 20, 1965. INEEL negative no. 65-241 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  10. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  11. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  12. Missiles caused by severe pressurized-water reactor accidients

    SciTech Connect

    Krieg, R.

    1995-07-01

    For future pressurized-water reactors, which should be designed against core-meltdown accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles, as well as their velocities, may vary to a great extent. Therefore a reliable proof of the containment integrity is very difficult. In this article the potential sources of missiles are discussed, and the conclusion was reached that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles that endanger the containment shell. Furthermore, during a melt-through failure of the reactor vessel under high pressure, the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions may be tolerable, but shielding structures that protect the containment shell may be required. Further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. Investigations are currently being carried out that will confirm the strength of the reactor vessel head. In addition, a device for retaining the fragments of a failing reactor vessel is discussed.

  13. Dynamic pressure approach to analysis of reactor fuel plate stability

    SciTech Connect

    Swinson, W.F.; Yahr, G.T.

    1990-01-01

    The dynamic pressure model can conveniently be used to evaluate the critical stress regions as a function of flow velocity. For some of the preliminary advanced neutron source reactor plate designs this could be very significant since the flow velocity could be limited by peak stresses in the plates more than by deflection or stability. The dynamic pressure results predicts the differential pressure across a plate as a function of flow velocity. The pressure differential can then be used to find the deflection and/or stress of the plate using traditional plate analyses. Instability would occur when plates are touching at mid-channel such that rapid oscillations of pressure can occur. The technique is conservative and gives a design limit for the plate. This model is one of several methods being used in the design of the ANS fuel elements. 10 refs., 4 figs.

  14. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  15. Anticipatory control of xenon in a pressurized water reactor

    SciTech Connect

    Impink, A.J. Jr.

    1987-02-10

    A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals, control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.

  16. Upper internals arrangement for a pressurized water reactor

    DOEpatents

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  17. The coolability limits of a reactor pressure vessel lower head

    SciTech Connect

    Theofanous, T.G.; Syri, S.

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  18. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S.

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  19. Process and apparatus for adding and removing particles from pressurized reactors

    DOEpatents

    Milligan, John D.

    1983-01-01

    A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

  20. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  1. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    SciTech Connect

    Brumovsky, M.; Polachova, H.

    1995-11-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber`s, Hardrath-Ohman`s as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared.

  2. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    SciTech Connect

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  3. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    SciTech Connect

    GRIFFIN, PATRICK J.

    1999-09-14

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation.

  4. Submerged bed versus unsaturated flow reactor: A pressurized hydrogenotrophic denitrification reactor as a case study.

    PubMed

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal

    2016-10-01

    The paper compares the main features of a submerged bed reactor (SuBR) with bubbling and recirculation of gas to those of an unsaturated flow reactor (uSFR) with liquid recirculation. A novel pressurized closed-headspace hydrogenotrophic denitrification system characterized by safe and economic utilization of H2 gas was used for the comparison. Under similar conditions, denitrification rates were lower in the SuBR as a result of a lower effective biofilm surface area and overall gas-liquid mass transfer coefficient kLa. Similar values of effluent DOC were achieved for both reactors, although effluent suspended solids concentration of the SuBR were substantially higher. On the other hand, the required cleaning frequency in the SuBR was 2.5 times lower. Moreover, the SuBR is expected to reduce the recirculation energy consumption by 0.35 kWh/m(3) treated. PMID:27424057

  5. Reactor pressure vessel head vents and methods of using the same

    SciTech Connect

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  6. Fabrication Flaws in Reactor Pressure Vessel Repair Welds

    SciTech Connect

    Schuster, George J.; Doctor, Steven R.

    2007-12-01

    This paper describes the fabrication flaw distribution and characterization in the repair weld metal of reactor pressure vessels. This work indicates that the large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the repair ends. Parametric analysis using an exponential fit is performed on the data. A description of repair flaw morphology is provided. Fabrication flaws in repairs are characterized using high sensitivity nondestructive ultrasonic testing, validation by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing.

  7. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  8. Low pressure stagnation flow reactor with a flow barrier

    DOEpatents

    Vosen, Steven R.

    2001-01-01

    A flow barrier disposed at the periphery of a workpiece for achieving uniform reaction across the surface of the workpiece, such as a semiconductor wafer, in a stagnation flow reactor operating under the conditions of a low pressure or low flow rate. The flow barrier is preferably in the shape of annulus and can include within the annular structure passages or flow channels for directing a secondary flow of gas substantially at the surface of a semiconductor workpiece. The flow barrier can be constructed of any material which is chemically inert to reactive gases flowing over the surface of the semiconductor workpiece.

  9. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    SciTech Connect

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  10. Research of a Supercritical Pressure Water Cooled Reactor in Korea

    SciTech Connect

    Bae, Yoon-Yeong; Joo, Hyung-Kook; Jang, Jinsung; Jeong, Yong-Hwan; Song, Jin-ho; Yoon, Han-Young; Yoo, Jung-Yul

    2004-07-01

    In this paper the activities on the supercritical pressure water-cooled reactor (SCWR) in Korea are briefly introduced. Four projects on a SCWR are being conducted in Korea. Three of them are supported by the I-NERI program while one is by KAERI. Two of the I-NERI-supported projects concern suitable materials for supercritical pressure and temperature, and radiation environment. The other I-NERI-supported project surveys numerically and experimentally the proper turbulence modeling for the numerical calculation of heat transfer phenomena at a supercritical condition. Heat transfer at a supercritical condition is being studied at KAERI experimentally using carbon dioxide as a coolant. The test loop is to be completed by the end of 2004. (authors)

  11. Behavior of stainless steels in pressurized water reactor primary circuits

    NASA Astrophysics Data System (ADS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-08-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  12. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  13. Biofilm architecture in a novel pressurized biofilm reactor.

    PubMed

    Jiang, Wei; Xia, Siqing; Duan, Liang; Hermanowicz, Slawomir W

    2015-01-01

    A novel pure-oxygen pressurized biofilm reactor was operated at different organic loading, mechanical shear and hydrodynamic conditions to understand the relationships between biofilm architecture and its operation. The ultimate goal was to improve the performance of the biofilm reactor. The biofilm was labeled with seven stains and observed with confocal laser scanning microscopy. Unusual biofilm architecture of a ribbon embedded between two surfaces with very few points of attachment was observed. As organic loading increased, the biofilm morphology changed from a moderately rough layer into a locally smoother biomass with significant bulging protuberances, although the chemical oxygen demand (COD) removal efficiency remained unchanged at about 75%. At higher organic loadings, biofilms contained a larger fraction of active cells distributed uniformly within a proteinaceous matrix with decreasing polysaccharide content. Higher hydrodynamic shear in combination with high organic loading resulted in the collapse of biofilm structure and a substantial decrease in reactor performance (a COD removal of 16%). Moreover, the important role of proteins for the spatial distribution of active cells was demonstrated quantitatively. PMID:25990377

  14. Iron Catalyst Chemistry in High Pressure Carbon Monoxide Nanotube Reactor

    NASA Technical Reports Server (NTRS)

    Scott, Carl D.; Povitsky, Alexander; Dateo, Christopher; Gokcen, Tahir; Smalley, Richard E.

    2001-01-01

    The high-pressure carbon monoxide (HiPco) technique for producing single wall carbon nanotubes (SWNT) is analyzed using a chemical reaction model coupled with properties calculated along streamlines. Streamline properties for mixing jets are calculated by the FLUENT code using the k-e turbulent model for pure carbon monixide. The HiPco process introduces cold iron pentacarbonyl diluted in CO, or alternatively nitrogen, at high pressure, ca. 30 atmospheres into a conical mixing zone. Hot CO is also introduced via three jets at angles with respect to the axis of the reactor. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Cluster reaction rates are from Krestinin, et aI., based on shock tube measurements. Another model is from classical cluster theory given by Girshick's team. The calculations are performed on streamlines that assume that a cold mixture of Fe(CO)5 in CO is introduced along the reactor axis. Then iron forms clusters that catalyze the formation of SWNTs from the Boudouard reaction on Fe-containing clusters by reaction with CO. To simulate the chemical process along streamlines that were calculated by the fluid dynamics code FLUENT, a time history of temperature and dilution are determined along streamlines. Alternative catalyst injection schemes are also evaluated.

  15. Flux effect analysis in WWER-440 reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Blagoeva, D.; Debarberis, L.

    2013-11-01

    The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

  16. Vitiated ethane oxidation in a high-pressure flow reactor

    SciTech Connect

    Walters, K.M.; Bowman, C.T.

    2009-10-15

    Vitiated combustion processes offer the potential to improve the thermodynamic efficiency in hydrocarbon-fueled combustion systems, providing a subsequent decrease in energy-specific CO{sub 2} emissions along with a decrease in the emission levels of nitrogen oxides (NO{sub x}) and particulate matter. The present work comprises an experimental and modeling study of vitiated ethane oxidation in a high-pressure flow reactor, with pressures of 1-6 bar, O{sub 2} mole fractions of 3.5-7.0%, temperatures of 1075-1100 K and 15-18 mole.% H{sub 2}O. Time-history measurements of species are used to characterize the overall rate of reaction and track the fuel-carbon through intermediate and product species. A one-dimensional mixing-reacting model that accounts for partial oxidation during reactant mixing is used in conjunction with a detailed kinetic mechanism. Changes in competing pathways due to variations in pressure and O{sub 2} mole fraction give rise to the complex pressure dependence seen in the experiments. (author)

  17. Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

    SciTech Connect

    Aleksakov, A. N. Yankovskiy, K. I.; Dunaev, V. I.; Kushbasov, A. N.

    2015-05-15

    The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

  18. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  19. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  20. Pressurized fluidized bed reactor and a method of operating the same

    DOEpatents

    Isaksson, J.

    1996-02-20

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  1. Pressurized fluidized bed reactor and a method of operating the same

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  2. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  3. Design of Recycle Pressurized Water Reactor with Heavy Water Moderation

    SciTech Connect

    Hibi, Koki; Uchita, Masato

    2004-03-15

    This study presents the conceptual design of the recycle pressurized water reactor (RPWR), which is an innovative PWR fueled with mixed oxide, moderated by heavy water, and having breeding ratios around 1.1. Most of the systems of RPWR can employ those of PWRs. The RPWR has no boric acid systems and has a small tritium removal system. The construction and operation costs would be similar to those of current PWRs. Heavy water cost has decreased drastically with up-to-date producing methods. The reliability of the systems of the RPWR is high, and the research and development cost for RPWR is very low because the core design is fundamentally based on the current PWR technology.

  4. Pressurized-water reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  5. Pressurized water reactor fuel crud and corrosion modeling

    NASA Astrophysics Data System (ADS)

    Deshon, Jeff; Hussey, Dennis; Kendrick, Brian; McGurk, John; Secker, Jeff; Short, Michael

    2011-08-01

    Pressurized water reactors circulate high-temperature water that slowly corrodes Inconel and stainless steel system surfaces, and the nickel/iron based corrosion products deposit in regions of the fuel where sub-cooled nucleate boiling occurs. The deposited corrosion products, called `crud', can have an adverse impact on fuel performance. Boron can concentrate within the crud in the boiling regions of the fuel leading to a phenomenon known as axial offset anomaly (AOA). In rare cases, fuel clad integrity can be compromised because of crud-induced localized corrosion (CILC) of the zirconium-based alloy. Westinghouse and the Electric Power Research Institute have committed to understanding the crud transport process and develop a risk assessment software tool called boron-induced offset anomaly (BOA) to avoid AOA and CILC. This paper reviews the history of the BOA model development and new efforts to develop a micro-scale model called MAMBA for use in the Consortium for Advanced Light Water Reactor Simulation (CASL) program.

  6. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    NASA Astrophysics Data System (ADS)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  7. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    SciTech Connect

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  8. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-13

    ... COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews... purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555...

  9. Russian Folktales.

    ERIC Educational Resources Information Center

    Field, Anne

    2000-01-01

    Discussion of Russian folktales suggests a structured approach to introducing the characters, settings, and motifs in an elementary classroom. Highlights include beginner-level texts; connections to Russian culture; writing connections; activities; and an annotated bibliography of 32 titles. (LRW)

  10. Embrittlement recovery due to annealing of reactor pressure vessel steels

    SciTech Connect

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1996-03-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  11. Improving CRDCS through digital technology for pressurized water reactors

    SciTech Connect

    Liddle, P.

    2006-07-01

    This paper describes increased Control Rod Drive Control System (CRDCS) reliability through the use of digital technology. A CRDCS is a non-safety plant system used to control rod motion into and out of the reactor core, which in turn controls the reactivity in the core. The CRDCS also provides the operator with information on the status of the control rods and the components within the system. The Digital CRDCS (DCRDCS) is an upgrade for the Babcock and Wilcox (B and W) pressurized water reactor control rod drive control system. The existing CRDCS was designed using relays, various electromechanical devices, and discrete solid-state components. The DCRDCS provides higher reliability, simplified maintenance, and streamlined system configuration through the use of Triple Modular Redundant (TMR) Controllers that implement the logic and control functions, as well as other design enhancements. A particularly important feature of the DCRDCS is the system's redundancy. With this configuration, no single failure of any DCRDCS component can cause an uncontrolled reaction. Such uncontrolled reactions might include withdrawal of the control rods, causing unwanted, or preventing, reactor trip, the loss of either or both the Absolute Position Indication (API) or Relative Position Indication (RPI), or inhibited rod movement. The DCRDCS is made up of two primary sections: the System Logic Equipment (SLE) and the Motor Control Equipment (MCE). The SLE is the primary component of the system. Signals transmitted to this equipment are processed by its logic, producing output signals that control the rod positions and provide system status information. The MCE receives command signals from the SLE to energize the Control Rod Drive Mechanism motors, producing the actual control rod motions. The DCRDCS is designed as a direct, state-of-the-art replacement for the existing CRDCS. Functionally, the DCRDCS will fulfill, and in some areas, exceed the functionality of the existing

  12. ARD remediation with limestone in a CO2 pressurized reactor

    USGS Publications Warehouse

    Sibrell, Philip L.; Watten, Barnaby J.; Friedrich, Andrew E.; Vinci, Brian J.

    2000-01-01

    We evaluated a new process for remediation of acid rock drainage (ARD). The process treats ARD with intermittently fluidized beds of granular limestone maintained within a continuous flow reactor pressurized with CO2. Tests were performed over a thirty day period at the Toby Creek mine drainage treatment plant, Elk County, Pennsylvania in cooperation with the Pennsylvania Department of Environmental Protection. Equipment performance was established at operating pressures of 0, 34, 82, and 117 kPa using an ARD flow of 227 L/min. The ARD had the following characteristics: pH, 3.1; temperature, 10 °C; dissolved oxygen, 6.4 mg/L; acidity, 260 mg/L; total iron, 21 mg/L; aluminum, 22 mg/L; manganese, 7.5 mg/L; and conductivity, 1400 μS/cm. In all cases tested, processed ARD was net alkaline with mean pH and alkalinities of 6.7 and 59 mg/L at a CO2 pressure of 0 kPa, 6.6 and 158 mg/L at 34 kPa, 7.4 and 240 mg/L at 82 kPa, and 7.4 and 290 mg/L at 117 kPa. Processed ARD alkalinities were correlated to the settled bed depth (p<0.001) and CO2 pressure (p<0.001). Iron, aluminum, and manganese removal efficiencies of 96%, 99%, and 5%, respectively, were achieved with filtration following treatment. No indications of metal hydroxide precipitation or armoring of the limestone were observed. The surplus alkalinity established at 82 kPa was successful in treating an equivalent of 1136 L/min (five-fold dilution) of the combined three ARD streams entering the Toby Creek Plant. This side-stream capability provides savings in treatment unit scale as well as flexibility in treatment effect. The capability of the system to handle higher influent acidity was tested by elevating the acidity to 5000 mg/L with sulfuric acid. Net alkaline effluent was produced, indicating applicability of the process to highly acidic ARD.

  13. The behavior of shallow flaws in reactor pressure vessels

    SciTech Connect

    Rolfe, S.T. )

    1991-11-01

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs.

  14. A New Scanning Tunneling Microscope Reactor Used for High Pressure and High Temperature Catalysis Studies

    SciTech Connect

    Tao, Feng; Tang, David C.; Salmeron, Miquel; Somorjai, Gabor A.

    2008-05-12

    We present the design and performance of a home-built high-pressure and high-temperature reactor equipped with a high-resolution scanning tunneling microscope (STM) for catalytic studies. In this design, the STM body, sample, and tip are placed in a small high pressure reactor ({approx}19 cm{sup 3}) located within an ultrahigh vacuum (UHV) chamber. A sealable port on the wall of the reactor separates the high pressure environment in the reactor from the vacuum environment of the STM chamber and permits sample transfer and tip change in UHV. A combination of a sample transfer arm, wobble stick, and sample load-lock system allows fast transfer of samples and tips between the preparation chamber, high pressure reactor, and ambient environment. This STM reactor can work as a batch or flowing reactor at a pressure range of 10{sup -13} to several bars and a temperature range of 300-700 K. Experiments performed on two samples both in vacuum and in high pressure conditions demonstrate the capability of in situ investigations of heterogeneous catalysis and surface chemistry at atomic resolution at a wide pressure range from UHV to a pressure higher than 1 atm.

  15. Advances in crack-arrest technology for reactor pressure vessels

    SciTech Connect

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  16. New techniques for modeling the reliability of reactor pressure vessels

    SciTech Connect

    Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.; Simonen, E.P.

    1986-01-01

    In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes several new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel wall thickness, and fluence distributions that vary throughout the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper. The effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithms for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RT/sub NDT/. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest toughness with subsequent initiation toughnesses.

  17. New techniques for modeling the reliability of reactor pressure vessels

    SciTech Connect

    Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.; Simonen, E.P.

    1985-12-01

    In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel walls thickness, and fluence distributions that vary through-out the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper. The effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithm for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RTNDT. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest thoughness with subsequent initiation toughnesses. 21 refs.

  18. THE DEVELOPMENT OF RADIATION EMBRITTLEMENT MODELS FOR U.S. POWER REACTOR PRESSURE VESSEL STEELS

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S

    2006-01-01

    The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  19. Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

    SciTech Connect

    Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.

    1993-08-01

    The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117.

  20. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    SciTech Connect

    Dickson, T.L.

    1993-04-01

    This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.

  1. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    SciTech Connect

    Dickson, T.L.

    1993-01-01

    This report discusses probabilistic fracture mechanics (PFM) analysis which is a major element of the comprehensive probabilistic methodology endorsed by the NRC for evaluation of the integrity of Pressurized Water Reactor (PWR) pressure vessels subjected to pressurized-thermal-shock (PTS) transients. It is anticipated that there will be an increasing need for an improved and validated PTS PFM code which is accepted by the NRC and utilities, as more plants approach the PTS screening criteria and are required to perform plant-specific analyses. The NRC funded Heavy Section Steel Technology (HSST) Program at Oak Ridge National Laboratories is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) PTS PFM code, which is intended to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as PFM global modeling methodology, the capability to approximate the effects of thermal streaming on circumferential flaws located inside a plume region created by fluid and thermal stratification, a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an adequate range of two and three dimensional inside surface flaws, the flexibility to generate a variety of output reports, and user friendliness.

  2. An idealized transient model for melt dispersal from reactor cavities during pressurized melt ejection accident scenarios

    SciTech Connect

    Tutu, N.K.

    1991-06-01

    The direct Containment Heating (DCH) calculations require that the transient rate at which the melt is ejected from the reactor cavity during hypothetical pressurized melt ejection accident scenarios be calculated. However, at present no models, that are able to predict the available melt dispersal data from small scale reactor cavity models, are available. In this report, a simple idealized model of the melt dispersal process within a reactor cavity during a pressurized melt ejection accident scenario is presented. The predictions from the model agree reasonably well with the integral data obtained from the melt dispersal experiments using a small scale model of the Surry reactor cavity. 17 refs., 15 figs.

  3. High-performance simulations for atmospheric pressure plasma reactor

    NASA Astrophysics Data System (ADS)

    Chugunov, Svyatoslav

    Plasma-assisted processing and deposition of materials is an important component of modern industrial applications, with plasma reactors sharing 30% to 40% of manufacturing steps in microelectronics production. Development of new flexible electronics increases demands for efficient high-throughput deposition methods and roll-to-roll processing of materials. The current work represents an attempt of practical design and numerical modeling of a plasma enhanced chemical vapor deposition system. The system utilizes plasma at standard pressure and temperature to activate a chemical precursor for protective coatings. A specially designed linear plasma head, that consists of two parallel plates with electrodes placed in the parallel arrangement, is used to resolve clogging issues of currently available commercial plasma heads, as well as to increase the flow-rate of the processed chemicals and to enhance the uniformity of the deposition. A test system is build and discussed in this work. In order to improve operating conditions of the setup and quality of the deposited material, we perform numerical modeling of the plasma system. The theoretical and numerical models presented in this work comprehensively describe plasma generation, recombination, and advection in a channel of arbitrary geometry. Number density of plasma species, their energy content, electric field, and rate parameters are accurately calculated and analyzed in this work. Some interesting engineering outcomes are discussed with a connection to the proposed setup. The numerical model is implemented with the help of high-performance parallel technique and evaluated at a cluster for parallel calculations. A typical performance increase, calculation speed-up, parallel fraction of the code and overall efficiency of the parallel implementation are discussed in details.

  4. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    SciTech Connect

    Spencer, Benjamin; Backman, Marie; Chakraborty, Pritam; Hoffman, William

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  5. Safety system augmentation at Russian nuclear power plants

    SciTech Connect

    Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.; Reister, R.A. |

    1996-12-31

    This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC power supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.

  6. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  7. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  8. Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

    SciTech Connect

    Bryson, J.W.; Dickson, T.L.; Malik, S.N.M.; Simonen, F.A.

    1999-08-01

    The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.

  9. Design and construction of a cascading pressure reactor prototype for solar-thermochemical hydrogen production

    NASA Astrophysics Data System (ADS)

    Ermanoski, Ivan; Grobbel, Johannes; Singh, Abhishek; Lapp, Justin; Brendelberger, Stefan; Roeb, Martin; Sattler, Christian; Whaley, Josh; McDaniel, Anthony; Siegel, Nathan P.

    2016-05-01

    Recent work regarding the efficiency maximization for solar thermochemical fuel production in two step cycles has led to the design of a new type of reactor—the cascading pressure reactor—in which the thermal reduction step of the cycle is completed in multiple stages, at successively lower pressures. This approach enables lower thermal reduction pressures than in single-staged reactors, and decreases required pump work, leading to increased solar to fuel efficiencies. Here we report on the design and construction of a prototype cascading pressure reactor and testing of some of the key components. We especially focus on the technical challenges particular to the design, and their solutions.

  10. Startup Thermal Considerations for Supercritical-Pressure Light Water-Cooled Reactors

    SciTech Connect

    Nakatsuka, Toru; Oka, Yoshiaki; Koshizuka, Seiichi

    2001-06-15

    Supercritical-pressure light water-cooled reactors (SCRs) are innovative systems aimed at high efficiency and cost reduction. The once-through direct-cycle plant system is the leading system of fossil-fired power plants (FPPs). Estimates of the coolability and necessary sizes of the SCR startup systems, sequences, and required equipment for startup are investigated with reference to supercritical FPPs. There are two types of supercritical boilers. One is a constant pressure boiler, and the other is a variable pressure boiler.First, startup of the constant pressure boiler is examined. The reactor starts at a supercritical pressure. A startup bypass system consisting of a flash tank and pressure-reducing valves is required. Second, startup of the variable pressure boiler is investigated. The reactor starts at a subcritical pressure, and the pressure increases with the load. A steam-water separator and a drain tank are required for startup.The results of computer calculations show that with both constant pressure and variable pressure startup, the peak cladding temperature does not exceed the operating limit through startup, and both startup sequences are feasible. The sizes of the components required for the startup systems are assessed. To simplify the plant system and to reduce the component size, variable pressure startup with steam separators in the bypass line appears desirable.

  11. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    SciTech Connect

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  12. Pressure-Letdown Machine for a Coal Reactor

    NASA Technical Reports Server (NTRS)

    Perkins, G. S.; Mabe, W. B.

    1986-01-01

    Pumps operating in reverse generate power. Conceptual pressure-letdown machine for coal-liquefaction system extracts energy from expansion of product fluid. Mud pumps, originally intended for use in oil drilling, operated in reverse so their motors act as generators. Several pumps operated in alternating phase to obtain multiple stages of letdown from inlet pressure to outlet pressure. About 75 percent of work generates inlet pressure recoverable as electrical energy.

  13. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    SciTech Connect

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L.

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  14. Advanced high-temperature, high-pressure transport reactor gasification

    SciTech Connect

    Swanson, M.L.

    1999-07-01

    The mission of the U.S. Department of Energy's (DOE's) Federal Energy Technology Center Office of Power Systems Product Management is to foster the development and deployment of advanced, clean, and affordable fossil-based (coal) power systems. These advanced power systems include the development and demonstration of gasification-based advanced power systems. These systems are integral parts of the Vision 21 Program for the co-production of power and chemicals which is being developed at DOE. DOE has been developing advanced gasification systems which lower the capital and operating cost of producing syngas for electricity or chemicals production. A transport reactor gasifier has shown potential to be a low-cost syngas producer as compared to other gasification systems because of its high throughput. This work directly supports the Power Systems Development Facility (PSDF) utilizing the Kellogg, Brown and Root (KBR) transport reactor located at the Southern Company Services (SCS) Wilsonville, Alabama, site. Over 1000 hours of operation on three different fuels in the pilot-scale transport reactor development unit (TRDU) has been completed to date. The Energy and Environmental Research Center (EERC) has established an extensive database on the operation of various fuels in a transport reactor gasifier. This database will be useful in determining the effectiveness of design changes on a transport reactor gasifier. It has been demonstrated that corrected fuel gas heating values ranging between 105 to 130 Btu/scf can be achieved. Factors that affect the TRDU product gas quality appear to be circulation rate, coal type, temperature, and air:coal and steam:coal ratios. Future plans are to modify the transport reactor mixing zone and J-leg loop seal to increase backmixing, thereby increasing solids residence time and gasifier performance. Enriched air- and oxygen-blown gasification tests, especially on widely available low-cost fuels such as petroleum coke, will also be

  15. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  16. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  17. Design, Construction and Operation Of A High Pressure Flow Loop Reactor For Carbon Sequestration

    SciTech Connect

    Gerdemann, Stephen J., Penner, Larry R.

    2003-11-01

    The Department of Energy’s Albany Research Center has been exploring the possibility of direct mineral carbonation as a means of sequestering carbon dioxide. As part of this research, a three-phase flow through reactor capable of operating at 200°C and 2500 psia was built. The reactor is a plug flow reactor with continuous and complete recycle. The results from this reactor may be used to design a larger and truly continuous flow reactor. This paper describes the design, construction and operation of this reactor. The extent of reaction, pressure drop across the pump and static mixers were measured at various test conditions. The extent of reaction was then compared to the results achievable in an autoclave.

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

    SciTech Connect

    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.

  19. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  20. Development and Application of the Reactor Coolant On-Line Leakage Evaluation Model for Pressurized Water Reactors

    SciTech Connect

    Liang, Thomas K.S.; Hung, H.-J.; Chang, C.-J.

    2001-12-15

    With the consideration of mass unbalance, coolant shrinking, and compressibility, a model for reactor coolant leakage evaluation has been developed to quantify on-line the system leakage rate with conventional system measurements, regardless of where the leak occurs. This model has been derived from the system of total continuity, and it divides the reactor coolant system (RCS) into two regions, namely, the saturated and subcooled regions. The pressurizer is considered as a saturated region, and the remaining part of the RCS is regarded as a subcooled region. Taking the on-line measurements of the RCS including the RCS pressure, temperature, pressurizer water level, and charging and letdown flow rates, this model can directly evaluate on-line the RCS leakage rate. It is noted that this model is applicable only if the RCS remains subcooled. To verify the applicability of this model, data generated by RELAP5/MOD3 simulation and experimental measurements from the Institute of Nuclear Energy Research, Taiwan, Integral System Test Facility were adopted to assess this model. With further on-line verification against the Maanshan training simulator, this model was finally delivered to the Maanshan nuclear power plant (a three-looped Westinghouse pressurized water reactor) to assist the operator training and on-line evaluation of the RCS leakage rate. The smallest amount of leak flow that can be detected by the ROCK model is 3 gal/min.

  1. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael Swanson; Daniel Laudal

    2008-03-31

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher

  2. Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization

    SciTech Connect

    1980-10-01

    This report summarizes work performed for the U.S. Arms Control and Disarmament Agency under BOA AC9NX707 (Task Order 80-02), as part of the Agency's continuing program on improved fuel utilization in light water reactors. The objective of the study was to investigate improvements in fuel management and design of water reactors (PWRs) that could potentially increase the utilization of natural uranium resources in a once-through fuel cycle (i.e., without using spent fuel reprocessing and recycle). For the present study, potential improvements were limited to retrofittable concepts, i.e., those which could be modifications to the reactor system or balance of plant. The potential improvements considered were not necessarily restricted to those which might be economical under current uranium ore prices or to those which might be acceptable to the nuclear industry at the present time. A six-month fuel cycle, for example, although technically possible, would be neither economical nor accept able to the industry at the present time. Although all potential improvements are not necessarily compatible with each other, the target objective was to seek a composite system of compatible improvements that, if possible, could increase uranium resource utilization by 30% or more. Economic factors, risks involved in the introduction, and potential licensing concerns are also addressed in the report.

  3. Pressure suppression containment system for boiling water reactor

    SciTech Connect

    Gluntz, D.M.; Nesbitt, L.B.

    1997-01-21

    A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

  4. Pressure suppression containment system for boiling water reactor

    DOEpatents

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  5. The Development of Radiation Embrittlement Models for U. S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

  6. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    SciTech Connect

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  7. Progress in understanding of direct containment heating phenomena in pressurized light water reactors

    SciTech Connect

    Ginsberg, T.; Tutu, N.K.

    1988-01-01

    Progress is described in development of a mechanistic understanding of direct containment heating phemonena arising during high-pressure melt ejection accidents in pressurized water reactor systems. The experimental data base is discussed which forms the basis for current assessments of containment pressure response using current lumped-parameter containment analysis methods. The deficiencies in available methods and supporting data base required to describe major phenomena occurring in the reactor cavity, intermediate subcompartments and containment dome are highlighted. Code calculation results presented in the literature are cited which demonstrate that the progress in understanding of DCH phenomena has also resulted in current predictions of containment pressure loadings which are significantly lower than are predicted by idealized, thermodynamic equilibrium calculations. Current methods are, nonetheless, still predicting containment-threatening loadings for large participating melt masses under high-pressure ejection conditions. Recommendations for future research are discussed. 36 refs., 5 figs., 1 tab.

  8. The first critical experiment with a LEU Russian fuel IRT-4M at the training reactor VR-1

    SciTech Connect

    Frybort, Jan

    2008-07-15

    A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replaced B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)

  9. Ab initio simulation of radiation damage in nuclear reactor pressure vessel materials

    NASA Astrophysics Data System (ADS)

    Watts, Daniel; Finkenstadt, Daniel

    2012-02-01

    Using Kinetic Monte Carlo we developed a code to study point defect hopping in BCC metallic alloys using energetics and attempt frequencies calculated using VASP, an electronic structure software package. Our code provides a way of simulating the effects of neutron radiation on potential reactor materials. Specifically we will compare the Molybdenum-Chromium alloy system to steel alloys for use in nuclear reactor pressure vessels.

  10. Optimization of Pressurized Oxy4Combustion with Flameless Reactor

    SciTech Connect

    Malavasi, Massimo; Landegger, Gregory

    2014-06-30

    Pressurized OxyECombustion is one of the most promising technologies for utilityEscale power generation plants. Benefits include the ability to burn low rank coal and capture C02. By increasing the flue gas pressure during this process, greater efficiencies are derived from increased quantity and quality of thermal energy recovery. UPA with modeling support from MIT and testing and data verification by Georgia Tech’s Research Center designed and built a 100kW system capable of demonstrating pressurized oxyEcombustion using a flameless combustor. Wyoming PRB coal was run at 15 and 32 bar. Additional tests were not completed but sampled data demonstrated the viability of the technology over a broader range of operating pressures, Modeling results illustrated a flat efficiency curve over 20 bar, with optimum efficiency achieved at 29 bar. This resulted in a 33% (HHV) efficiency, a 5 points increase in efficiency versus atmospheric oxyEcombustion, and a competitive cost of electricity plus greater C02 avoidance costs then prior study’s presented. UPA’s operation of the benchEscale system provided evidence that key performance targets were achieved: flue gas sampled at the combustor outlet had nonE detectable residual fly ashes, and low levels of SO3 and heavyEmetal. These results correspond to prior pressurized oxyEcombustion testing completed by IteaEEnel.

  11. Component failures at pressurized water reactors. Final report

    SciTech Connect

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  12. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  13. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  14. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  15. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, Paul R.

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  16. Apparatus and process to eliminate diffusional limitations in a membrane biological reactor by pressure cycling

    DOEpatents

    Efthymiou, George S.; Shuler, Michael L.

    1989-08-29

    An improved multilayer continuous biological membrane reactor and a process to eliminate diffusional limitations in membrane reactors in achieved by causing a convective flux of nutrient to move into and out of an immobilized biocatalyst cell layer. In a pressure cycled mode, by increasing and decreasing the pressure in the respective layers, the differential pressure between the gaseous layer and the nutrient layer is alternately changed from positive to negative. The intermittent change in pressure differential accelerates the transfer of nutrient from the nutrient layers to the biocatalyst cell layer, the transfer of product from the cell layer to the nutrient layer and the transfer of byproduct gas from the cell layer to the gaseous layer. Such intermittent cycling substantially eliminates mass transfer gradients in diffusion inhibited systems and greatly increases product yield and throughput in both inhibited and noninhibited systems.

  17. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  18. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  19. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    SciTech Connect

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  20. Combining COMSOL modeling with acoustic pressure maps to design sono-reactors.

    PubMed

    Wei, Zongsu; Weavers, Linda K

    2016-07-01

    Scaled-up and economically viable sonochemical systems are critical for increased use of ultrasound in environmental and chemical processing applications. In this study, computational simulations and acoustic pressure maps were used to design a larger-scale sono-reactor containing a multi-stepped ultrasonic horn. Simulations in COMSOL Multiphysics showed ultrasonic waves emitted from the horn neck and tip, generating multiple regions of high acoustic pressure. The volume of these regions surrounding the horn neck were larger compared with those below the horn tip. The simulated acoustic field was verified by acoustic pressure contour maps generated from hydrophone measurements in a plexiglass box filled with water. These acoustic pressure contour maps revealed an asymmetric and discrete distribution of acoustic pressure due to acoustic cavitation, wave interaction, and water movement by ultrasonic irradiation. The acoustic pressure contour maps were consistent with simulation results in terms of the effective scale of cavitation zones (∼ 10 cm and <5 cm above and below horn tip, respectively). With the mapped acoustic field and identified cavitation location, a cylindrically-shaped sono-reactor with a conical bottom was designed to evaluate the treatment capacity (∼ 5 L) for the multi-stepped horn using COMSOL simulations. In this study, verification of simulation results with experiments demonstrates that coupling of COMSOL simulations with hydrophone measurements is a simple, effective and reliable scientific method to evaluate reactor designs of ultrasonic systems. PMID:26964976

  1. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  2. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    SciTech Connect

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  3. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    SciTech Connect

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  4. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J.; Seren, T.; Lipponen, M.; Kekki, T.

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  5. Design and performance of a high-pressure Fischer-Tropsch fluidized bed reactor

    SciTech Connect

    Weimer, A.W.; Quarderer, G.J.; Cochran, G.A.; Conway, M.M. )

    1988-01-01

    A 900 kg/day, CO/H/sub 2/, high-pressure, fluidized bed, pilot reactor was designed from first principles to achieve high reactant conversions and heat removal rates for the Fischer-Tropsch (F-T) synthesis of liquefied petroleum gases (LPG's). Suppressed bubble growth at high pressure allowed high reactant conversions which nearly matched those obtained at identical conditions in a lab scale fixed bed reactor. For GHSV approximately 1400 hr/sup -1/ and T = 658 {Kappa} at P approximately 7000 {kappa}Pa, reactant conversion exceeded 75%. The reactor heat removal capability exceeded twice design performance with the fluidized bed easily operating under thermally stable conditions. The fluidized catalyst was a potassium promoted, molybdenum on carbon (Mo/{Kappa}/C) catalyst which did not produce any detrimental waxy products. Long catalyst lifetimes of 1000 hrs on steam between regenerations allowed the fluidized bed to be operated in a batch mode.

  6. A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

    SciTech Connect

    Beach, N. J.; Knapp, S. M. M.; Landis, C. R.

    2015-10-15

    The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from −90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor to be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.

  7. Advanced Computational Thermal Studies and their Assessment for Supercritical-Pressure Reactors (SCRs)

    SciTech Connect

    D. M. McEligot; J. Y. Yoo; J. S. Lee; S. T. Ro; E. Lurien; S. O. Park; R. H. Pletcher; B. L. Smith; P. Vukoslavcevic; J. M. Wallace

    2009-04-01

    The goal of this laboratory / university collaboration of coupled computational and experimental studies is the improvement of predictive methods for supercritical-pressure reactors. The general objective is to develop supporting knowledge needed of advanced computational techniques for the technology development of the concepts and their safety systems.

  8. Hydrolysis of Tifton 85 bermudagrass in a pressurized batch hot water reactor

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Background: Ethanol production from grass is desirable due to the large amount of biomass it produces. However, a pretreatment is necessary before fermentation to increase ethanol yield. Tifton 85 bermudagrass was treated with a newly designed pressurized batch hot water reactor. Multiple temperatur...

  9. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-19

    ... supplementing a notice published in the Federal Register on March 20, 2012 (77 FR 16270), that requested public...; email: Evelyn.Gettys@nrc.gov . SUPPLEMENTARY INFORMATION: On March 20, 2012 (77 FR 16270), the NRC... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized...

  10. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Requirements for thermal annealing of the reactor pressure vessel. 50.66 Section 50.66 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... Program. The percent recovery of RTNDT and Charpy upper-shelf energy due to the thermal...

  11. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Requirements for thermal annealing of the reactor pressure vessel. 50.66 Section 50.66 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... Program. The percent recovery of RTNDT and Charpy upper-shelf energy due to the thermal...

  12. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Requirements for thermal annealing of the reactor pressure vessel. 50.66 Section 50.66 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... Program. The percent recovery of RTNDT and Charpy upper-shelf energy due to the thermal...

  13. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Requirements for thermal annealing of the reactor pressure vessel. 50.66 Section 50.66 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... Program. The percent recovery of RTNDT and Charpy upper-shelf energy due to the thermal...

  14. Stochastic model to monitor mechanical vibrations in pressurized water reactors

    SciTech Connect

    Shieh, D.J.; Upadhyaya, B.R.

    1984-01-01

    The feasibility of using neutron flux and core-exit temperature signals in PWRs for estimating core coolant flow velocity has been demonstrated using normal operational data from both the LOFT reactor and a commerical PWR. The LOFT analysis further showed that the core coolant velocity can be accurately monitored for various flow rates using the linear phase-frequency relationship in the frequency range 0.1 to 2 Hz. The development of the technique for monitoring core coolant velocity in PWRs provides a valuable alternative for flow measurement. Theoretical studies of core heat transfer in PWRs showed that the fluctuating heat sources have a dominating effect on the core-exit temperature compared to fluctuations of the coolant flow rate and core inlet coolant temperature. In the present analysis a detailed distributed parameter model of a PWR core was developed with the purpose of studying the following aspects of core coolant flow rate measurement: the mechanisms causing linear phase relationship between neutron flux and coolant temperature signals due to various perturbation sources; the effect of axial flux shape on the phase slope (or estimated transit delay time); and the relationship between transit delay time and effective distance of temperature noise propagation to maintain the flow velocity invariant.

  15. Depletion optimization of lumped burnable poisons in pressurized water reactors

    SciTech Connect

    Kodah, Z.H.

    1982-01-01

    Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration.

  16. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

    The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

  17. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    SciTech Connect

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  18. The Information Fusion Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    The complex nonlinear dependencies observed in typical reactor pressure vessel (RPV) material embrittlement data, as well as the inherent large uncertainties and scatter in the radiation embrittlement data, make prediction of radiation embrittlement a difficult task. Conventional statistical and deterministic approaches have only resulted in rather large uncertainties, in part because they do not fully exploit domain-specific mechanisms. The domain models built by researchers in the field, on the other hand, do not fully exploit the statistical and information content of the data. As evidenced in previous studies, it is unlikely that a single method, whether statistical, nonlinear, or domain model, will outperform all others. More generally, considering the complexity of the embrittlement prediction problem, it is highly unlikely that a single best method exists and is tractable, even in theory. In this paper, we propose to combine a number of complementary methods including domain models, neural networks, and nearest neighbor regressions (NNRs). Such a combination of methods has become possible because of recent developments in measurement-based optimal fusers in the area of information fusion. The information fusion technique is used to develop radiation embrittlement prediction models for reactor RPV steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six Cu, Ni, P, neutron fluence, irradiation time, and irradiation-parameters are used in the embrittlement prediction models. The results-temperature indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The

  19. Modelling of NO destruction in a low-pressure reactor by an Ar plasma jet: species abundances in the reactor

    NASA Astrophysics Data System (ADS)

    Kutasi, Kinga

    2011-03-01

    The destruction of NO molecules by an Ar plasma jet in a low-pressure (0.2 Torr) reactor is investigated by means of a 3D hydrodynamic model. The density distribution of species created through molecular kinetics triggered by the collision of Ar+ with NO is calculated, showing that in the case of the most abundant species a quasi-homogeneous density distribution builds up in a large part of the reactor. The conversion of NO into stable O2 and N2 molecules is followed under different plasma jet conditions and NO gas flows, and the effect of N2 addition on NO destruction is studied. It is shown that in the present system the reproduction of NO molecules on the surface through surface-assisted recombination of N and O atoms becomes impossible due to the fast disappearance of N atoms in the jet's inlet vicinity.

  20. Autogenerative high pressure digestion: anaerobic digestion and biogas upgrading in a single step reactor system.

    PubMed

    Lindeboom, R E F; Fermoso, F G; Weijma, J; Zagt, K; van Lier, J B

    2011-01-01

    Conventional anaerobic digestion is a widely applied technology to produce biogas from organic wastes and residues. The biogas calorific value depends on the CH, content which generally ranges between 55 and 65%. Biogas upgrading to so-called 'green gas', with natural gas quality, generally proceeds with add-on technologies, applicable only for biogas flows > 100 m3/h. In the concept of autogenerative high pressure digestion (AHPD), methanogenic biomass builds up pressure inside the reactor. Since CO2 has a higher solubility than CH4, it will proportion more to the liquid phase at higher pressures. Therefore, AHPD biogas is characterised by a high CH4 content, reaching equilibrium values between 90 and 95% at a pressure of 3-90 bar. In addition, also H2S and NH3 are theoretically more soluble in the bulk liquid than CO2. Moreover, the water content of the already compressed biogas is calculated to have a dew point <--10 degrees C. Ideally, high-quality biogas can be directly used for electricity and heat generation, or injected in a local natural gas distribution net. In the present study, using sodium acetate as substrate and anaerobic granular sludge as inoculum, batch-fed reactors showed a pressure increase up to 90 bars, the maximum allowable value for our used reactors. However, the specific methanogenic activity (SMA) of the sludge decreased on average by 30% compared to digestion at ambient pressure (1 bar). Other results show no effect of pressure exposure on the SMA assessed under atmospheric conditions. These first results show that the proposed AHPD process is a highly promising technology for anaerobic digestion and biogas upgrading in a single step reactor system. PMID:22097043

  1. I and C modernization for VVER reactors. [Eastern bloc pressurized water reactors

    SciTech Connect

    Gangloff, W.C.; Werner, C.L. )

    1993-08-01

    Many countries of the former Eastern Bloc are now looking westward for the introduction, absorption, and widespread application of modern technology in many areas. One such area is the nuclear power industry, where an infusion of western technology is desired to improve both plant production economics and long term plant safety. Modern digital I and C systems are sought for upgrading the existing Soviet designed reactor plants to levels of safety and reliability consistent with western standards. This paper describes the functional and physical design of the distributed microprocessor based protection, control and plant information systems to be used in Czechoslovakia for the modernization of the Temelin units (VVER-1,000). The modernization incorporates not only new hardware, but also a new functional design based on a safety analysis performed to western standards. The new systems will reduce the maintenance requirements and costs and provide such enhancements as online testability and automated surveillance testing. The technology is directly applicable to both new reactor designs under construction and to retrofits for the older reactors. It is expected to find many applications in eastern and central Europe as well as in the former Soviet Union for improving the safety and reliability of their vital nuclear power stations.

  2. Ultrasound pressure distributions generated by high frequency transducers in large reactors.

    PubMed

    Leong, Thomas; Coventry, Michael; Swiergon, Piotr; Knoerzer, Kai; Juliano, Pablo

    2015-11-01

    The performance of an ultrasound reactor chamber relies on the sound pressure level achieved throughout the system. The active volume of a high frequency ultrasound chamber can be determined by the sound pressure penetration and distribution provided by the transducers. This work evaluated the sound pressure levels and uniformity achieved in water by selected commercial scale high frequency plate transducers without and with reflector plates. Sound pressure produced by ultrasonic plate transducers vertically operating at frequencies of 400 kHz (120 W) and 2 MHz (128 W) was characterized with hydrophones in a 2 m long chamber and their effective operating distance across the chamber's vertical cross section was determined. The 2 MHz transducer produced the highest pressure amplitude near the transducer surface, with a sharp decline of approximately 40% of the sound pressure occurring in the range between 55 and 155 mm from the transducer. The placement of a reflector plate 500 mm from the surface of the transducer was shown to improve the sound pressure uniformity of 2 MHz ultrasound. Ultrasound at 400 kHz was found to penetrate the fluid up to 2 m without significant losses. Furthermore, 400 kHz ultrasound generated a more uniform sound pressure distribution regardless of the presence or absence of a reflector plate. The choice of the transducer distance to the opposite reactor wall therefore depends on the transducer plate frequency selected. Based on pressure measurements in water, large scale 400 kHz reactor designs can consider larger transducer distance to opposite wall and larger active cross-section, and therefore can reach higher volumes than when using 2 MHz transducer plates. PMID:26186816

  3. Design Strategies for Optically-Accessible, High-Temperature, High-Pressure Reactor

    SciTech Connect

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  4. Design strategies for optically-accessible, high-temperature, high-pressure reactor

    SciTech Connect

    S. F. Rice; R. R. Steeper; C. A. LaJeunesse; R. G. Hanush; J. D. Aiken

    2000-02-01

    The authors have developed two optical cell designs for high-pressure and high-temperature fluid research: one for flow systems, and the other for larger batch systems. The flow system design uses spring washers to balance the unequal thermal expansions of the reactor and the window materials. A typical design calculation is presented showing the relationship between system pressure, operating temperature, and torque applied to the window-retaining nut. The second design employs a different strategy more appropriate for larger windows. This design uses two seals: one for the window that benefits from system pressure, and a second one that relies on knife-edge, metal-to-metal contact.

  5. Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI Programs

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1994-02-01

    This report discusses development on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels containing flaws. Fracture mechanics tests on reactor pressure vessel steel have shown that local brittle zones do not significantly degrade the material fracture toughness, constraint relaxation at the crack tip of shallow surface flaws results in increased fracture toughness, and biaxial loading reduces but does not eliminate the shallow-flaw fracture toughness elevation. Experimental irradiation investigations have shown that the irradiation-induced shift in Charpy V-notch versus temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement and the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  6. Response of biodegradation characteristics of unacclimated activated sludge to moderate pressure in a batch reactor.

    PubMed

    Xu, Rui-Xiao; Li, Bing; Zhang, Yong; Si, Ling; Zhang, Xian-Qiu; Xie, Biao

    2016-04-01

    This study was aimed to investigate the effect of moderate pressure on unacclimated activated sludge. Process of organic degradation, variation of carbon dioxide (CO2) concentration of off-gas and characteristics of extracellular polymeric substances (EPS) of activated sludge were analyzed using pressure-atmospheric comparative experiments in bench-scale batch reactors. It was found that moderate pressure increased the degradation rate more dramatically when the biological process ran under a higher organic load with much more oxygen demand, which illuminated that applications of the pressurized method to high concentration organic wastewaters would be more reasonable and practicable. High oxygen transfer impetus increased utilization of oxygen which not only promoted the biodegradation of organics in wastewater, but also led to more EPS consumption in activated sludge. CO2 concentration of off-gas was lower in the earlier stage due to CO2 being pressed into the liquid phase and converted into inorganic carbon (IC). More CO2 emission was observed during the pressurized aerobic process 160 min later. EPS in pressurized reactor was much lower, which may be an important way of sludge reduction by pressurized technology. PMID:26802261

  7. VISA: a computer code for predicting the probability of reactor pressure-vessel failure. [PWR

    SciTech Connect

    Stevens, D.L.; Simonen, F.A.; Strosnider, J. Jr.; Klecker, R.W.; Engel, D.W.; Johnson, K.I.

    1983-09-01

    The VISA (Vessel Integrity Simulation Analysis) code was developed as part of the NRC staff evaluation of pressurized thermal shock. VISA uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics are used to model crack initiation and propagation. parameters for initial crack size, copper content, initial RT/sub NDT/, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents the version of VISA used in the NRC staff report (Policy Issue from J.W. Dircks to NRC Commissioners, Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982, SECY-82-465) and includes a user's guide for the code.

  8. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  9. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  10. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    SciTech Connect

    Kulesza, J.A.; Fero, A.H.; Rouden, J.; Green, E.L.

    2011-07-01

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

  11. A novel Y-type reactor for selective excitation of atmospheric pressure glow discharge plasma

    NASA Astrophysics Data System (ADS)

    Xia, Guan-Guang; Wang, Jin-Yun; Huang, Aimin; Suib, Steven L.; Hayashi, Yuji; Matsumoto, Hiroshige

    2001-02-01

    A novel Y-type atmospheric pressure ac glow discharge plasma reactor has been designed and tested in CO reduction with hydrogen and the reverse water-gas shift reaction. The reactor consists of a Y-type quartz tube with an angle of 120°-180° between the two long arms, two metal rod electrodes serving as high voltage terminals and two pieces of aluminum foil which were wrapped outside of the quartz tubes as a ground electrode. Different combinations of input power applied on this three- electrode system can lead to selective plasmas on one side, two sides, or can also generate a stable arc between the two high voltage terminal electrodes. The ability to selectively activate different species with this type of apparatus can help to minimize side reactions in plasmas to obtain desirable products. The Y-type reactor may provide a novel means to study fundamental problems regarding radical reactions.

  12. Flaw density examinations of a clad boiling water reactor pressure vessel segment

    SciTech Connect

    Cook, K.V.; McClung, R.W.

    1986-01-01

    Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel (nominally 0.7 by 3 m (2 by 10 ft)). This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected.

  13. Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Bury, Tomasz

    2011-12-01

    Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

  14. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    SciTech Connect

    Brumovsky, M.; Steele, L.E.

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  15. Effect of long-term thermal aging on magnetic property in reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Sato, H.; Iwawaki, T.; Yamamoto, T.; Klingensmith, D.; Odette, G. R.; Kikuchi, H.; Kamada, Y.

    2013-08-01

    Effect of long-term thermal aging at 290 and 500 °C on magnetic hysteresis property in reactor pressure vessel steels and simple model alloys have been investigated for times up to 8800 h. While Vickers hardness is insensitive to thermal aging at both temperatures, coercivity generally exhibits a slight decrease after aging at 290 °C. In particular, at a higher temperature of 500 °C a steady increase of coercivity was observed for reactor pressure vessel steels, whereas coercivity for simple model alloys exhibits an abrupt drop just after aging and the decrease was 20-30% of that before aging. The results were interpreted by the thermally-assisted formation of Cu-rich precipitates and recovery, but the latter has the dominant effect for simple model alloys because of their ferritic microstructure. The possible effect of relaxation of lattice strain created by dissolved interstitial atoms during neutron irradiation is proposed.

  16. The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

    SciTech Connect

    Dickson, T.L.; Bass, B.R.; McAfee, W.J.

    1998-01-01

    Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K{sub Ic}, was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4.

  17. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  18. A study on applicability of decay ratio estimation in a pressurized water reactor

    SciTech Connect

    Por, G. . Central Research Inst. for Physics); Runkel, J. . Nuclear Engineering and Nondestructing Testing Inst.)

    1994-03-01

    Decay ratio was estimated via a simplified method from the impulse response function that had been evaluated using an unvariable autoregression method. Suggested estimation was utilized in neutron noise measurements carried out during seven fuel cycles of a 1,300-MW (electric) pressurized water reactor. Results show that such an evaluation method can be used to monitor the increasing oscillation of the neutron flux during the fuel cycle.

  19. A Reactor Pressure Vessel Dosimetry Calculation Using ATTILA, An Unstructured Tetrahedral Mesh Discrete-Ordinates Code

    SciTech Connect

    Wareing, T.A.; Parsons, D.K.; Pautz, S.

    1997-12-31

    Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. In this paper we describe the application of ATTILA to a 3-D reactor pressure vessel dosimetry problem. We provide numerical results from ATTILA and the Monte Carlo code, MCNP. The results demonstrate the effectiveness and efficiency of ATTILA for such calculations.

  20. Computational fluid dynamic analysis of a closure head penetration in a pressurized water reactor

    SciTech Connect

    Forsyth, D.R.; Schwirian, R.E.

    1995-09-01

    ALLOY 600 has been used typically for penetrations through the closure head in pressurized water reactors because of its thermal compatibility with carbon steel, superior resistance to chloride attack and higher strength than the austenitic stainless steels. Recent plant operating experience with this alloy has indicated that this material may be susceptible to degradation. One of the major parameters relating to degradation of the head penetrations are the operational temperatures and stress levels in the penetration.

  1. Advanced Pressurized Water Reactor for Improved Resource Utilization: Part I - Survey of Potential Improvements

    SciTech Connect

    Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W. III

    1981-09-15

    This document is an interim report under ACDA BOA AC9NX707, Task Order 80-03, which covers the evaluation of certain potential improvements in pressurized water reactor designs intended to enhance uranium fuel utilization. The objective of these evaluations is to seek advanced, non-retrofittable improvements that could possibly be commercialized by the end of the century, and, on the basis of a preliminary evaluation, to select compatible improvements for incorporation into a composite advanced pressurized water reactor concept. The principal areas of investigation include reduced parasitic absorption of neutrons (Task 1), reduced neutron leakage (Task 2), and alternative fuel design concepts (Task 3). To the extent possible, the advanced concept developed in an earlier study (Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization, SSA-128, October 1980) is used as a basis in developing the advanced composite concept. The reference design considered typical of present PWR commercial practice is the system described in RESAR-414, Reference Safety Analysis Report, Westinghouse Nuclear Energy Systems, October 1976.

  2. Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Almazouzi, A.

    2009-03-01

    The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.

  3. Validation of medium-pressure UV disinfection reactors by Lagrangian actinometry using dyed microspheres.

    PubMed

    Shen, C; Scheible, O K; Chan, P; Mofidi, A; Yun, T I; Lee, C C; Blatchley, E R

    2009-03-01

    Lagrangian actinometry (LA) has been demonstrated to represent an alternative to conventional biodosimetry for validation of ultraviolet (UV) disinfection systems used in drinking water treatment. However, previous applications of LA for this purpose have all involved monochromatic (lambda = 254 nm) UV reactor systems. To address this issue, dyed microspheres (DMS) were applied for quantification of dose distribution delivery by field-scale UV reactor systems based on medium-pressure Hg lamp (MP) technology. These MP reactor systems are characterized by polychromatic output. Dose distribution estimates developed by LA for these reactors were reported as equivalent 254 nm distributions. When combined with the UV(254) dose-response behavior for challenge organisms used in simultaneous or parallel biodosimetry experiments, the dose distribution estimates developed from the microspheres yielded estimates of challenge organism inactivation that were in agreement with measured values. For one of the reactors tested, biodosimetry tests were conducted with two challenge organisms that had different UV dose-response behavior; UV dose distribution estimates from LA yielded predictions of microbial inactivation that were in agreement with measured inactivation responses for both challenge organisms for all test conditions. It is likely that the agreement between LA results and biodosimetry data was related, in part, to the agreement between the action spectra of the microspheres and the challenge organisms. Because LA yields a measure of the UV dose distribution delivered by a reactor, the information from LA assays will eliminate many sources of uncertainty in the design and operation of UV systems, thereby allowing for implementation of UV reactor systems that are less expensive than their predecessors, yet more reliable. PMID:19138781

  4. Laser anemometry measurements of natural circulation flow in a scale model PWR reactor system. [Pressurized Water Reactor

    NASA Technical Reports Server (NTRS)

    Kadambi, J. R.; Schneider, S. J.; Stewart, W. A.

    1986-01-01

    The natural circulation of a single phase fluid in a scale model of a pressurized water reactor system during a postulated grade core accident is analyzed. The fluids utilized were water and SF6. The design of the reactor model and the similitude requirements are described. Four LDA tests were conducted: water with 28 kW of heat in the simulated core, with and without the participation of simulated steam generators; water with 28 kW of heat in the simulated core, with the participation of simulated steam generators and with cold upflow of 12 lbm/min from the lower plenum; and SF6 with 0.9 kW of heat in the simulated core and without the participation of the simulated steam generators. For the water tests, the velocity of the water in the center of the core increases with vertical height and continues to increase in the upper plenum. For SF6, it is observed that the velocities are an order of magnitude higher than those of water; however, the velocity patterns are similar.

  5. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    NASA Astrophysics Data System (ADS)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  6. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  7. Advanced Computational Modeling of Vapor Deposition in a High-pressure Reactor

    NASA Technical Reports Server (NTRS)

    Cardelino, Beatriz H.; Moore, Craig E.; McCall, Sonya D.; Cardelino, Carlos A.; Dietz, Nikolaus; Bachmann, Klaus

    2004-01-01

    In search of novel approaches to produce new materials for electro-optic technologies, advances have been achieved in the development of computer models for vapor deposition reactors in space. Numerical simulations are invaluable tools for costly and difficult processes, such as those experiments designed for high pressures and microgravity conditions. Indium nitride is a candidate compound for high-speed laser and photo diodes for optical communication system, as well as for semiconductor lasers operating into the blue and ultraviolet regions. But InN and other nitride compounds exhibit large thermal decomposition at its optimum growth temperature. In addition, epitaxy at lower temperatures and subatmospheric pressures incorporates indium droplets into the InN films. However, surface stabilization data indicate that InN could be grown at 900 K in high nitrogen pressures, and microgravity could provide laminar flow conditions. Numerical models for chemical vapor deposition have been developed, coupling complex chemical kinetics with fluid dynamic properties.

  8. Advanced Computational Modeling of Vapor Deposition in a High-Pressure Reactor

    NASA Technical Reports Server (NTRS)

    Cardelino, Beatriz H.; Moore, Craig E.; McCall, Sonya D.; Cardelino, Carlos A.; Dietz, Nikolaus; Bachmann, Klaus

    2004-01-01

    In search of novel approaches to produce new materials for electro-optic technologies, advances have been achieved in the development of computer models for vapor deposition reactors in space. Numerical simulations are invaluable tools for costly and difficult processes, such as those experiments designed for high pressures and microgravity conditions. Indium nitride is a candidate compound for high-speed laser and photo diodes for optical communication system, as well as for semiconductor lasers operating into the blue and ultraviolet regions. But InN and other nitride compounds exhibit large thermal decomposition at its optimum growth temperature. In addition, epitaxy at lower temperatures and subatmospheric pressures incorporates indium droplets into the InN films. However, surface stabilization data indicate that InN could be grown at 900 K in high nitrogen pressures, and microgravity could provide laminar flow conditions. Numerical models for chemical vapor deposition have been developed, coupling complex chemical kinetics with fluid dynamic properties.

  9. Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

    SciTech Connect

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

    The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

  10. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  11. Numerical simulation of turbulent flow in the throttle of the MBIR reactor's low-pressure chamber

    NASA Astrophysics Data System (ADS)

    Yarunichev, V. A.; Orlova, E. E.; Lemekhov, Yu. V.; Shpanskii, V. A.

    2015-08-01

    This work in devoted to numerical calculation of turbulent flow in a labyrinth-type throttle. A system of such throttles is installed at the inlet to the MBIR reactor's low-pressure chamber and serves for setting up the required pressure difference and coolant flow rate. MBIR is a multipurpose fourthgeneration fast-neutron research reactor intended for investigating new kinds of nuclear fuel, structural materials, and coolants. The aim of this work is to develop a verified procedure for carrying out 3D calculation of the throttle using CFD modeling techniques. The investigations on determining the throttle hydraulic friction coefficient were carried out in the range of Reynolds numbers Re = 52000-136000. The reactor coolant (liquid sodium) was modeled by tap water. The calculations were carried out using high-Reynolds-number turbulence models with the near-wall functions k-ɛ and RNG k-ɛ, where k is the turbulent pulsation kinetic energy and ɛ is the turbulence kinetic energy dissipation rate. The obtained results have shown that the calculated value of hydraulic friction coefficient differs from its experimental value by no more than 10%. The developed procedure can be applied in determining the hydraulic friction coefficient of a modified labyrinth throttle design. The use of such calculation will make it possible to predict an experiment with the preset accuracy.

  12. Conceptual design of a pressure tube light water reactor with variable moderator control

    SciTech Connect

    Rachamin, R.; Fridman, E.; Galperin, A.

    2012-07-01

    This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

  13. Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

    SciTech Connect

    Lu, S.C.; Sommer, S.C.; Johnson, G.L. ); Lambert, H.E. )

    1990-10-01

    This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns.

  14. Effects of temperature and pressure on the in-reactor creepdown of Zircaloy fuel cladding

    SciTech Connect

    Hobson, D.O.; Thoms, K.R.; Dodd, C.V.; van der Kaa, Th.

    1982-01-01

    Descriptions and results for seven of the eight in-reactor creepdown tests of Zircaloy fuel cladding, which were part of a joint program between the U.S. Nuclear Regulatory Commission and Energieonderzoek Centrum Nederland, are presented. These tests were conducted to study the behavior of Zircaloy fuel cladding under conditions that approximate those found in an operating pressurized-water power reactor. The most important conclusion to be drawn from this study involves the deformation of the cladding during testing. Contrary to similar tests conducted out-of-reactor, the in-reactor specimens did not deform uniformly, that is, by diametral contraction and smooth ovalization. Rather, the deformation surfaces were nonuniform with hills and valleys being formed at irregular intervals. This implies that conventional concepts of creep rate and simplified modeling procedures will not work for predicting cladding behavior. Sufficient data have been generated in this program to supply modelers with detailed descriptions of the cladding surface shapes from which new interpretations can be derived to predict cladding behavior.

  15. Evaluation of HFIR (High Flux Isotope Reactor) pressure-vessel integrity considering radiation embrittlement

    SciTech Connect

    Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

    1988-04-01

    The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of approx.10/sup 4/ less), that is, a rate effect.

  16. Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels

    SciTech Connect

    Hawthorne, J.R.; Hiser, A.L. )

    1990-03-01

    This report describes a set of experiments undertaken using a 2 MW test reactor, the UBR, to qualify the significance of fluence rate to the extent of embrittlement produced in reactor pressure vessel steels at their service temperature. The test materials included two reference plates (A 302-B, A 533-B steel) and two submerged arc weld deposits (Linde 80, Linde 0091 welding fluxes). Charpy-V (C{sub v}), tension and 0.5T-CT compact specimens were employed for notch ductility, strength and fracture toughness (J-R curve) determinations, respectively. Target fluence rates were 8 {times} 10{sup 10}, 6 {times} 10{sup 11} and 9 {times} 10{sup 12} n/cm{sup 2} {minus}s{sup {minus}1}. Specimen fluences ranged from 0.5 to 3.8 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The data describe a fluence-rate effect which may extend to power reactor surveillance as well as test reactor facilities now in use. The dependence of embrittlement sensitivity on fluence rate appears to differ for plate and weld deposit materials. Relatively good agreement in fluence-rate effects definition was observed among the three test methods. 52 figs., 4 tabs.

  17. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    SciTech Connect

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  18. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    NASA Astrophysics Data System (ADS)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  19. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    DOE PAGESBeta

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

  20. Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

    NASA Astrophysics Data System (ADS)

    Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

    2003-06-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. The measurements were made using Red Perspex™ polymethyl methacrylate high-level film dosimeters, a Radcal "peanut" ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

  1. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    SciTech Connect

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.

  2. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, Roy C.; Gou, Perng-Fei; Chu, Cherk Lam; Oliver, Robert P.

    1999-01-01

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

  3. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, R.C.; Gou, P.F.; Chu, C.L.; Oliver, R.P.

    1999-07-27

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block. 6 figs.

  4. Influence of crack depth on the fracture toughness of reactor pressure vessel steel

    SciTech Connect

    Theiss, T.J.; Bryson, J.W.

    1991-01-01

    The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the crack tip. The loss of constraint takes place when interaction occurs between the elastic-plastic crack-tip stress field and the specimen surface nearest the crack tip. An increased shallow-crack fracture toughness is of interest to the nuclear industry because probabilistic fracture-mechanics evaluations show that shallow flaws play a dominant role in the probability of vessel failure during postulated pressurized-thermal-shock (PTS) events. Tests have been performed on beam specimens loaded in 3-point bending using unirradiated reactor pressure vessel material (A533 B). Testing has been conducted using specimens with a constant beam depth (W = 94 mm) and within the lower transition region of the toughness curve for A533 B. Test results indicate a significantly higher fracture toughness associated with the shallow flaw specimens compared to the fracture toughness determined using deep-crack (a/W = 0.5) specimens. Test data also show little influence of thickness on the fracture toughness for the current test temperature ({minus}60{degree}C). 21 refs., 5 figs., 3 tabs.

  5. Russians as People.

    ERIC Educational Resources Information Center

    Miller, Wright

    This analysis of the Russian character in various aspects of Soviet society in its daily activities focuses on the cultural rather than the political. Included in the study are sections on: (1) hibernation and awakening; (2) the Russian scene; (3) being a Russian; (4) Russian society--mass and minority; (5) manners, morals, and taste; and (6)…

  6. Detection and characterization of flaws in segments of light water reactor pressure vessels

    SciTech Connect

    Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

    1987-01-01

    Studies have been conducted to determine flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessil and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication (with a through-wall dimension of approx.6 mm (approx.0.24 in.)) was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications (i.e., for a total of approximately 6.8 m/sup 2/ (72 ft/sup 2/) of cladding surface).

  7. Transit time of mixed high pressure injection water and primary loop water in pressurized water reactor cold legs

    SciTech Connect

    Sun, B.H.; Oh, S.; Rothe, P.H.

    1984-03-01

    During an overcooling transient in a pressurized water reactor, cold water from the high pressure injection (HPI) mixes with the hot primary coolant in the cold leg. The transit time is a gauge for the assessment of the time and the velocity of the mixed flow that passes through the cold leg to the downcomer. Existing data from mixing tests at the Electric Power Research Institute (EPRI)/CREARE and EPRI/SAI facilities are analyzed. By means of models for HPI jet entrainment as well as the propagation of a gravity current, dimensionless correlations have been developed for the transit time and cold water front velocity at stagnant loop flow conditions. Based on this transit time correlation for stagnant loop flow and the limiting condition for large loop flow, a general correlation has been developed to account for the loop flow effect on transit time. These correlations unify a wide range of data obtained from five geometrically different test sections with two fluids (pure water and saline solution). In addition to the geometric factors, the governing dimensionless parameters for the transit time are the HPI jet Froude number, the Froude number for the cold-leg channel, and the ratio of loop flow to HPI flow.

  8. Pressure drop considerations of a lithium cooled fusion breeder tokamak reactor blanket

    SciTech Connect

    Wong, C.P.C.

    1983-12-06

    Liquid lithium was selected as one of the coolants for the 1983 fusion breeder blanket used on the magnetically confined tokamak fusion reactor, and as a result, the thermal-hydraulic calculations were dominated by magnetohydrodynamic (MHD) considerations. The applicable sets of MHD equations for the engineering thermal-hydraulic design were reviewed and compared. Special attention was given to the MHD calculations for the fertile material zone, a packed bed of composite beryllium and thorium balls, since this region can dominate the thermal-hydraulic behavior of this blanket module. To keep the pressure drops acceptable, fertile fuel balls were omitted in the inboard blanket.

  9. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    SciTech Connect

    Le Pape, Yann; Huang, Hai

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

  10. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    NASA Astrophysics Data System (ADS)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  11. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    SciTech Connect

    Al-Falahi, A.; Haennine, M.; Porkholm, K.

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  12. Atmospheric pressure chemical vapor deposition of CdTe—reactor design considerations

    NASA Astrophysics Data System (ADS)

    Meyers, Peter V.; Kee, Robert J.; Raja, Laxminarayan; Wolden, Colin A.; Aire, Michael

    1999-03-01

    Atmospheric Pressure Chemical Vapor Deposition (APCVD) of polycrystalline thin-film CdTe appears to offer several practical advantages over state-of-the-art manufacturing techniques. APCVD employs the same reaction chemistry utilized to produce 16% efficient CdTe cells (i.e., same reaction chemistry as Close Spaced Sublimation), avoids use of vacuum equipment, allows for physical separation of the source and substrate, and employs forced convection to ensure uniform delivery of source material over large-area substrates. Reactor design considerations and preliminary numerical simulations of mass transport are presented.

  13. MECHANICAL PROPERTIES AND MICROSTRUCTURE OF THREE RUSSIAN Mechanical Properties And Microstructure Of Three Russian Ferritic/Martensitic Steels Irradiated In BN-350 Reactor To 50 dpa at 490C

    SciTech Connect

    Dvoriashin, Alexander M; Porollo, S I; Konobeev, Yu V; Budylkin, N I; Minonova, E G; Loltukhovsky, A G; Leonteva-Smirnova, M V; Bochvar, A A; Garner, Francis A

    2007-03-01

    Ferritic/martensitic (F/M) steels are being considered for application in fusion reactors, intense neutron sources, and accelerator-driven systems. While EP-450 is traditionally used with sodium coolants in Russia, EP-823 and EI-852 steels with higher silicon levels have been developed for reactor facilities using lead-bismuth coolant. To determine the influence of silicon additions on short-term mechanical properties and microstructure, ring specimens cut from cladding tubes of these three steels were irradiated in sodium at 490°С in the BN-350 reactor to 50 dpa. Post-irradiation tensile testing and microstructural examination show that EI-852 steel (1.9 wt% Si) undergoes severe irradiation embrittlement. Microstructural investigation showed that the formation of near-continuous phase precipitates on grain boundaries is the main cause of the embrittlement.

  14. The effects of plasma inhomogeneity on the nanoparticle coating in a low pressure plasma reactor

    SciTech Connect

    Pourali, N.; Foroutan, G.

    2015-10-15

    A self-consistent model is used to study the surface coating of a collection of charged nanoparticles trapped in the sheath region of a low pressure plasma reactor. The model consists of multi-fluid plasma sheath module, including nanoparticle dynamics, as well as the surface deposition and particle heating modules. The simulation results show that the mean particle radius increases with time and the nanoparticle size distribution is broadened. The mean radius is a linear function of time, while the variance exhibits a quadratic dependence. The broadening in size distribution is attributed to the spatial inhomogeneity of the deposition rate which in turn depends on the plasma inhomogeneity. The spatial inhomogeneity of the ions has strong impact on the broadening of the size distribution, as the ions contribute both in the nanoparticle charging and in direct film deposition. The distribution width also increases with increasing of the pressure, gas temperature, and the ambient temperature gradient.

  15. Detection of small-sized near-surface under-clad cracks for reactor pressure vessels

    SciTech Connect

    Taylor, T.T.; Crawford, S.L.; Doctor, S.R.; Posakony, G.J.

    1983-02-01

    The analysis of pressurized thermal shock (PTS) shows it is necessary for nondestructive evaluation to demonstrate high probability of detecting evaluation to demonstrate high probability of detecting cracks 0.250 inches deep and deeper at the clad/base metal interface. Ultrasonic techniques developed and used in Europe are evaluated in this paper for their applicability to US reactor pressure vessels for detecting cracks of interest for PTS. Flaw detectability experiments were carried out by testing the inspection technique's ability to detect artificial flaws under several types of clad, including some Manual Metal Arc (MMA) clad. Both ground and unground clad surfaces were evaluated. Crack sizing tests of the inspection technique were made using a crack tip diffraction technique.

  16. Estimation of the uncertainty in TRAC/PF1-MOD1 predictions of production reactor plenum pressures

    SciTech Connect

    Griggs, D.P. )

    1992-01-01

    The TRAC-PF1/MOD1 code (TRAC) is used to perform best-estimate analyses of certain postulated design-basis accidents (DBAs) in Savannah River Site (SRS) production reactors. One of the DBAs analyzed is an instantaneous double-ended guillotine break loss-of-coolant accident (LOCA). The TRAC analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in a detailed analysis of a single fuel assembly. The quantification of uncertainty is an important element in determining safe operating power levels for SRS reactors. This motivates the estimation of the uncertainty in using spatial interpolations of the relatively coarse cell-average plenum pressure predictions obtained with TRAC to predict detailed reactor plenum pressure distributions. This result supports the adequacy of the {plus minus}5% plenum pressure uncertainty estimated for LOCA analyses.

  17. In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

    SciTech Connect

    Sweeney, F.J.; Upadhyaya, B.R.; Shieh, D.J.

    1984-01-01

    Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements. Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.

  18. Analysis of in-core dynamics in pressurized water reactors with application to parameter monitoring

    SciTech Connect

    Upadhyaya, B.R.; Sweeney, F.J.; Shieh, D.J.; Glockler, O.

    1987-01-01

    The behavior of the phase relationship between neutron flux and core-exit temperature fluctuations in a pressurized water reactor (PWR) is studied as a function of the moderator temperature coefficient of reactivity (..cap alpha../sub c/). PWR operational data indicate that the neutron noise and core-exit temperature noise cross power spectrum phase is linear in a certain frequency range, and approaches -180 deg at low frequencies. Extensive modeling studies applied to the LOFT reactor show that this low frequency phase behavior changes when ..cap alpha../sub c/ is positive, approaching zero deg at low frequencies. The analysis further showed that in the LOFT reactor, coolant flow rate fluctuation is the primary driving source causing neutron noise and core-exit temperature fluctuations. This conclusion was also confirmed by independent studies. The neutron noise-coolant temperature phase behavior may be used as a single method of monitoring the moderator temperature coefficient of reactivity during different stages of a PWR fuel cycle.

  19. A microfluidic reactor for rapid, low-pressure proteolysis with on-chip electrospray ionization.

    PubMed

    Liuni, Peter; Rob, Tamanna; Wilson, Derek J

    2010-02-01

    A microfluidic reactor that enables rapid digestion of proteins prior to on-line analysis by electrospray ionization mass spectrometry (ESI-MS) is introduced. The device incorporates a wide (1.5 cm), shallow (10 microm) reactor 'well' that is functionalized with pepsin-agarose, a design that facilitates low-pressure operation and high clogging resistance. Electrospray ionization is carried out directly from a short metal capillary integrated into the chip outlet. Fabrication, involving laser ablation of polymethyl methacrylate (PMMA), is exceedingly straightforward and inexpensive. High sequence coverage spectra of myoglobin (Mb), ubiquitin (Ub) and bovine serum albumin (BSA) digests were obtained after <4 s of residence time in the reactor. Stress testing showed little loss of performance over approximately 2 h continuous use at high flow rates (30 microL/min). The device provides a convenient platform for a range of applications in proteomics and structural biology, i.e. to enable high-throughput workflows or to limit back-exchange in spatially resolved hydrogen/deuterium exchange (HDX) experiments. PMID:20049884

  20. Determination of core design thermal safety limits for a two-loop pressurized water reactor

    SciTech Connect

    Kostadinov, V.

    1996-04-01

    Results are given of independent research of core thermal design limits for the Nuklearna Elektrarna Krsko (NEK) nuclear power plant; procedures for two-loop pressurized water reactor plant core design safety limit calculation are used. Emphasis is placed on researching the vessel exit boiling and the hot-channel exit quality limits and their impact on the maximum available design safety operating range and thermal operating margin of the NEK reactor core. For this purpose, the LIMITS computer code is developed. Based on the modified, well-tried COBRA-IV-I computer code, the departure of nuclear boiling ratio core safety limits are calculated. The original results complement well those of the NEK Final Safety Analysis Report. The procedures and the methods for determining the reactor core design thermal limits are successfully proven despite the unavailability of proprietary data, different models, and computer codes. In addition to the acquired capability of in-house independent checking of the vendor`s results, the bases are set for further independent analyses of the limiting safety system settings for the NEK core.

  1. Back corona enhanced organic film deposition inside an Atmospheric Pressure Weakly Ionized Plasma reactor

    NASA Astrophysics Data System (ADS)

    Islam, Rokibul; Xie, Shuzheng; Englund, Karl; Pedrow, Patrick

    2014-10-01

    A grounded screen with short needle-like protrusions has been designed to generate back corona in an Atmospheric Pressure Weakly Ionized Plasma (APWIP) reactor. The grounded screen with protrusions is placed downstream at a variable gap length from an array of needles that is energized with 60 Hz high voltage. The excitation voltage is in the range 0--10 kV RMS and the feed gas mixture consists of argon and acetylene. A Lecroy 9350AL 500 MHz digital oscilloscope is used to monitor the reactor voltage and current using a resistive voltage divider and a current viewing resistor, respectively. The current signal contains many positive and negative current pulses associated with corona discharge. Analysis of the current signal shows asymmetry between positive and negative corona discharge currents. Photographs show substantial back corona generated near the tips of the protrusions situated at the grounded screen. The back corona activates via bond scission acetylene radicals that are transported downstream to form a plasma-polymerized film on a substrate positioned downstream from the grounded screen. The oscillograms will be used to generate corona mode maps that show the nature of the corona discharge as a function of gap spacing, applied voltage and many other reactor parameters.

  2. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    SciTech Connect

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  3. On the thermal stability of late blooming phases in reactor pressure vessel steels: An atomistic study

    NASA Astrophysics Data System (ADS)

    Bonny, G.; Terentyev, D.; Bakaev, A.; Zhurkin, E. E.; Hou, M.; Van Neck, D.; Malerba, L.

    2013-11-01

    Radiation-induced embrittlement of bainitic steels is the lifetime limiting factor of reactor pressure vessels in existing nuclear light water reactors. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. In view of improving the predictive capability of existing models it is necessary to understand better the mechanisms leading to the formation of these defects, amongst which the so-called "late blooming phases". In this work we study the stability of the latter by means of density functional theory (DFT) calculations and Monte Carlo simulations based on a here developed quaternary FeCuNiMn interatomic potential. The potential is based on extensive DFT and experimental data. The reference DFT data on solute-solute interaction reveal that, while Mn-Ni pairs and triplets are unstable, larger clusters are kept together by attractive binding energy. The NiMnCu synergy is found to increase the temperature range of stability of solute atom precipitates in Fe significantly as compared to binary FeNi and FeMn alloys. This allows for thermodynamically stable phases close to reactor temperature, the range of stability being, however, very sensitive to composition.

  4. Implementation of an expert system for xenon spatial control in pressurized-water reactors

    SciTech Connect

    Chung, S.K.

    1988-01-01

    Control of the axial xenon oscillations is a knowledge- and experience-intensive activity for reactor operators. To aid reactor operators in the control of axial xenon oscillations, an advisory expert system was developed. A rule-based expert system shell, INSIGHT2+, was used to build the expert system which was interfaced with a microcomputer-based core control model of a pressurized-water reactor, graphic engine, and data base. A core control model described by one-group diffusion theory with moderator temperature and xenon feedbacks was used to develop heuristic control rules and to test the system. Full- and part-length control rods, boron concentration, and coolant inlet temperature were considered as control variables of the core control model. This expert system consists of a search space: the set of possible power level and power shape patterns. The search space was made by combining the following core state variables: the sign of relative power and axial offset (AO) error, sign of the rate of change of power level and AO, and magnitude of relative power and AO error.

  5. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    SciTech Connect

    Schulz, K.C.; Yahr, G.T.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  6. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    SciTech Connect

    Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

    2014-02-18

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

  7. Global analysis of bundle behavior in pressurized water reactor specific CORA experiments

    SciTech Connect

    Hering, W. ); Minato, Kazuo; Nagase, Fumihisa )

    1993-04-01

    At Kernforschungszentrum Karlsruhe, out-of-pile bundle experiments are performed in the CORA facility to investigate the behavior of light water reactor fuel elements during severe fuel damage accidents. To analyze the phenomena observed during the tests, such as claddin failure, oxidation, and deformation, as well as their influence on the post test bundle state, four pressurized water reactor specific tests are selected: CORA-2, CORA-3, CORA-5, and CORA-12. From each of these tests, a detailed global analysis using all the measured temperatures, pressures, and fluid compositions as well as videoscope information has been performed. To describe the post test bundle state quantitatively, axial profiles of the bundle cross-section area, the damage state of the rods, the average cladding oxidation, and the damage to the pellets are measured. The effects of CORA-specific components on the bundle melt progression and the measured axial profiles are identified and assessed. Most of the observations during the tests as well as the post test bundle state can be explained by the established common sequence of phenomena. For a better understanding of the melt progression, some physical phenomena, such as the energy release associated with the double-sided oxidation of the cladding, the melt release, or the melt relocation, must be analyzed in detail.

  8. Passive safety injection experiments with a large-scale pressurized water reactor simulator

    SciTech Connect

    Yonomoto, Taisuke; Kukita, Yutaka; Anoda, Yoshinari; Asaka, Hideaki

    1995-03-01

    Two experiments were conducted at the ROSA-V/ Large-Scale-Test-Facility to investigate thermal-hydraulic behavior of a gravity-driven passive injection system for a pressurized water reactor under cold-leg small break loss-of-coolant accident conditions. The injection system, used in the tests, consisted of a tank located above the reactor vessel, an injection line, and pressure balance lines. The two tests were conducted using the same break area, corresponding to 2.5% of the scaled cold-leg cross-sectional area, and different actuation logic for the automatic depressurization system (ADS). Both experimental results showed an accumulation of hot water in the upper part of the tank due to the natural circulation, followed by a continuous water level drop, and the existence of a slightly superheated liquid layer near the water surface. Because of the differences in the ADS actuation logic, the system depressurization behavior was different between the two tests. Much larger injection rates from the tank were obtained for the test that experienced the larger depressurization rate. The liquid temperature distributions obtained from these tests were predicted well by an analytical model proposed in a previous paper.

  9. VISA-II: a computer code for predicting the probability of reactor pressure vessel failure

    SciTech Connect

    Simonen, F.A.; Johnson, K.I.; Liebetrau, A.M.; Engel, D.W.; Simonen, E.P.

    1986-03-01

    The VISA-II (Vessel Integrity Simulation Analysis code was originally developed as part of the NRC staff evaluation of pressurized thermal shock. VISA-II uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics methods are used to model crack initiation and propagation. Parameters for initial crack size and location, copper content, initial reference temperature of the nil-ductility transition, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents an upgraded version of the original VISA code as described in NUREG/CR-3384. Improvements include a treatment of cladding effects, a more general simulation of flaw size, shape and location, a simulation of inservice inspection, an updated simulation of the reference temperature of the nil-ductility transition, and treatment of vessels with multiple welds and initial flaws. The code has been extensively tested and verified and is written in FORTRAN for ease of installation on different computers. 38 refs., 25 figs.

  10. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy. Between 2004 and 2006, four fuel segments were irradiated, with on-line recording of centerline temperature and rod pressure of the two instrumented rods and intermittent non-destructive hot-cell investigations of the other two non-instrumented rods. At the end of 2006, the instrumented rods were unloaded for hot-cell investigations. The hot-cell investigations reduced uncertainties in the power history to build a reliable and consistent irradiation history which can be used to assess and validate fuel performance codes. The on-line recorded temperatures of the instrumented rods are presented in this paper and are compared to corresponding calculations on the basis of the power history. One of the non-instrumented rods was re-inserted in the reactor in 2012 and attained a peak burnup level of 37 GWd/tHM by the end of 2014. The combined data set of on-line measurements and post irradiation examinations enables further code validation. In this context, the results of the in-house MACROS code of SCK·CEN have been compared with the experimental results. The code contains dedicated (Th,Pu)O2 models for the calculation of the thermal conductivity as a function of the burnup and models that determine the radial power profile within the pellet.

  11. Partial Defect Verification of the Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2010-02-05

    The International Atomic Energy Agency (IAEA) has the responsibility to carry out independent inspections of all nuclear material and facilities subject to safeguards agreements in order to verify compliance with non-proliferation commitments. New technologies have been continuously explored by the IAEA and Member States to improve the verification measures to account for declared inventory of nuclear material and detect clandestine diversion and production of nuclear materials. Even with these efforts, a technical safeguards challenge has remained for decades for the case of developing a method in identifying possible diversion of nuclear fuel pins from the Light Water Reactor (LWR) spent fuel assemblies. We had embarked on this challenging task and successfully developed a novel methodology in detecting partial removal of fuel from pressurized water reactor spent fuel assemblies. The methodology uses multiple tiny neutron and gamma detectors in the form of a cluster and a high precision driving system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly without any movement of the fuel. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. The combined information of gamma and neutron signature is used to produce base signatures and they are principally dependent on the geometry of the detector locations, and exhibit little sensitivity to initial enrichment, burn-up or cooling time. A small variation in the fuel bundle such as a few missing pins changes the shape of the signature to enable detection. This resulted in a breakthrough method which can be used to detect pin diversion without relying on the nuclear power plant operator's declared operation data. Presented are the results of various Monte Carlo simulation studies and experiments from actual commercial PWR spent fuel assemblies.

  12. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    SciTech Connect

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.

  13. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  14. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 3-Surry Unit 1 Cycle 2

    SciTech Connect

    Bowman, S.M.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using selected critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations in this report is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of two reactor critical configurations for Surry Unit 1 Cycle 2. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted a direct comparison of criticality calculations using the utility-calculated isotopics with those using the isotopics generated by the SCALE-4

  15. Russian Translation.

    PubMed

    O'dette, R E

    1957-03-29

    This discussion has described the status of the large United States program for translation from the Russian. A partial description of what is being done or planned, and by whom, has been provided as a guide for those who wish to follow the subject further. The urge to pass on useful information has necessarily restricted the space which might also have been profitably devoted to the philosophic aspects of the problem. Although it is not said with any sense of pride in achievement-because much more remains to be done than has been done-it would seem fair to describe the current national translation activity, including all contributions to it, as a phenomenon. Phenomena in scientific communication are not common: a full appreciation of their significance requires more analysis than results from a simple listing of their outward characteristics. But a few observations might be made in conclusion. Most United States scientists probably feel that, as a nation, we are and should be world leaders in science, even though this feeling is neither nurtured nor expressed in a spirit of violent competition. If this assumption is allowed, the point which seems to remain is that the United States will not retain its position casually. Our scientists expect to maintain an awareness of the scientific achievements and failures of the other nations of the world. But we must especially become more aware of the advances of Soviet science, both qualitatively and quantitatively. The evidence points toward this last conclusion, regardless of whether one is concerned with the production of ideas or things, increase in man's knowledge of himself and his environment, conflict between idealisms, or simply the national security. PMID:17836422

  16. Microstructure and mechanical properties of austenitic stainless steel 12X18H9T after neutron irradiation in the pressure vessel of BR-10 fast reactor at very low dose rates

    NASA Astrophysics Data System (ADS)

    Porollo, S. I.; Dvoriashin, A. M.; Konobeev, Yu. V.; Ivanov, A. A.; Shulepin, S. V.; Garner, F. A.

    2006-12-01

    Results are presented for void swelling, microstructure and mechanical properties of Russian 12X18H9T (0.12C-18Cr-9Ni-Ti) austenitic stainless steel irradiated as a pressure vessel structural material of the BR-10 fast reactor at ˜350 °C to only 0.64 dpa, produced by many years of exposure at the very low displacement rate of only 1.9 × 10 -9 dpa/s. In agreement with a number of other recent studies it appears that lower dpa rates have a pronounced effect on the microstructure and resultant mechanical properties. In general, lower dpa rates lead to the onset of swelling at much lower doses compared to comparable irradiations conducted at higher dpa rates.

  17. Microstructure and mechanical properties of austenitic stainless steel 12X18H9T after neutron irradiation in the pressure vessel of BR-10 fast reactor at very low dose rates

    SciTech Connect

    Porollo, S. I.; Dvoriashin, Alexander M.; Konobeev, Yury V.; Ivanov, A. A.; Shulepin, S. V.; Garner, Francis A.

    2006-12-01

    Results are presented for void swelling, microstructure andmechanical properties of Russian 12X18H9T (0.12C-18Cr-9Ni-Ti) austenitic stainless steel irradiated as a pressure vessel structure material of the BR-10 fast reactor at ~350C to only 0.64 dpa, produced by many years of exposure at the very low displacement rate of only 1.9x10-9 dpa/s. In agreement with a number of other recent studies it appears that lower dpa rates have a pronounced effect on the microstructure and resultant mechanical properties. In general, loweer dpa rates lead to the onset of swelling at much lower doses compared to comparable irradiations conducted at higher dpa rates.

  18. Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

    SciTech Connect

    Balkey, K.; Witt, F.J.; Bishop, B.A.

    1995-06-01

    Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industry efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.

  19. European Pressurized water Reactor (EPR) SAR ATWS Accident Analyses by using 3D Code Internal Coupling Method

    SciTech Connect

    Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.

    2004-07-01

    Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)

  20. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  1. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    SciTech Connect

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  2. Boric acid corrosion of light water reactor pressure vessel head materials.

    SciTech Connect

    Park, J.-H.; Chopra, O. K.; Natesan, K.; Shack, W. J.; Cullen, Jr.; W. H.; Energy Technology; USNRC

    2005-01-01

    This work presents experimental data on electrochemical potential and corrosion rates for the materials found in the reactor pressure vessel head and control rod drive mechanism (CRDM) nozzles in boric acid solutions of varying concentrations at temperatures of 95-316 C. Tests were conducted in (a) high-temperature, high-pressure aqueous solutions with a range of boric acid concentrations, (b) high-temperature (150-316 C)H-B-Osolutions at ambient pressure, in wet and dry conditions, and (c) low-temperature (95 C) saturated, aqueous, boric acid solutions. These correspond to the following situations: (a) low leakage through the nozzle and nozzle/head annulus plugged, (b) low leakage through the nozzle and nozzle/head annulus open, and (c) significant cooling due to high leakage and nozzle/head annulus open. The results showed significant corrosion only for the low-alloy steel and no corrosion for Alloy 600 or 308 stainless steel cladding. Also, corrosion rates were significant in saturated boric acid solutions, and no material loss was observed in H-B-O solution in the absence of moisture. The results are compared with the existing corrosion/wastage data in the literature.

  3. Abatement characteristics of N2O in low-pressure plasma reactor

    NASA Astrophysics Data System (ADS)

    Hur, M.; Lee, J. O.; Lee, J. Y.; Kang, W. S.; Song, Y.-H.

    2016-02-01

    The abatement characteristics of N2O were investigated in a plasma reactor positioned in front of a vacuum pump. Fourier transform infrared (FTIR) spectroscopy was used for evaluating the destruction and removal efficiency (DRE) and the identification of byproducts. The concentration of NO x (NO and NO2) was quantified by using an NO x analyzer. The DRE of N2O was enhanced by increasing the power or decreasing the N2O flow rate. A higher pressure yields a higher DRE of N2O and a lower concentration of NO x in the destroyed N2O. For understanding this phenomenon, the discharge characteristics were analyzed by using optical emission spectroscopy (OES). The spatial distribution of emission spectra from the discharge in a mixture of N2O and Ar gases was measured by varying the pressure. The mechanisms underlying the pressure effect on the N2O DRE and NO x concentration are discussed in terms of the electron density and the concentration of O radicals.

  4. Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

    PubMed

    Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

    2004-08-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions. PMID:15220719

  5. Radwaste generation survey update: Volume 2, Pressurized water reactors: Final report

    SciTech Connect

    Daloisio, G.S.; Deltete, C.P.

    1988-02-01

    The Electric Power Research Institute (EPRI) commissioned an operations-related project (RP1557-26) in mid-1986 to update the project data base developed for EPRI Report NP-3370, ''Identification of Radwaste Sources and Reduction Techniques,'' which was published in January 1984. An update was deemed particularly desirable in order to assess the impact on power reactor low level radioactive waste generation of 10 CFR 61, the recent implementation of the 1985 Amendment to the Low Level Waste Policy Act of 1980 (and its potential effects on accelerated waste shipment programs), and the efforts of several plants to implement waste minimization program over the past several years. These events, as reflected in waste generation rates from 1982 through 1985, should help NP-3370 continue to be a useful document for a plant's radwaste manager in the future. Furthermore, the trends of the past several years presented herein should help to more accurately define utility waste source terms for use in planning on-site storage and developing regional burial facilities. A new data base was developed that includes 1982 through 1986 information, as well as pertinent portions of the 1978 through 1981 data base. The result of the project is a two volume report comprising radwaste related information from more than 95% of the nuclear power plants in commerical operation as of 1986. Volume 1 contains all information pertaining to boiling water reactors (BWRs), while Volume 2 contains information for pressurized water reactors (PWRs). The computerized data base of waste volumes, sources and characteristics for each plant type (BWR or PWR) is included as an appendix in each respective volume. 36 figs., 26 tabs.

  6. Pressure Drop Characteristics in Tight-Lattice Bundles for Reduced-Moderation Water Reactors

    NASA Astrophysics Data System (ADS)

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4×5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier.

  7. Notch position in the HAZ specimen of reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Kim, J. H.; Yoon, E. P.

    1998-12-01

    Variations in the notch toughness in the heat-affected zone (HAZ) were investigated by positioning the Charpy V-notches along the line normal to the weld fusion line of a SA 508 Cl.3 reactor pressure vessel (RPV) steel. In the notch position for common surveillance HAZ specimens, rather higher toughness values were acquired. The minimum properties were noted in the region of 4-5 mm apart from the fusion boundary, where the values of toughness and strength were both poorer than those of the other regions of the HAZ and the base metal. The causes for these variations were discussed with reference to the microstructures from the actual and the simulated welding processes.

  8. Development of a shallow-flaw fracture assessment methodology for nuclear reactor pressure vessels

    SciTech Connect

    Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Pennell, W.E.

    1996-06-01

    Shallow-flaw fracture technology is being developed within the Heavy-Section Steel Technology (HSST) Program for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVs) containing postulated shallow flaws. Cleavage fracture in shallow-flaw cruciform beam specimens tested under biaxial loading at temperatures in the lower transition temperature range was shown to be strain-controlled. A strain-based dual-parameter fracture toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture. A probabilistic fracture mechanics (PFM) model that includes both the properties of the inner-surface stainless-steel cladding and a biaxial shallow-flaw fracture toughness correlation gave a reduction in probability of cleavage initiation of more than two orders of magnitude from an ASME-based reference case.

  9. A Unified Cohesive Zone Approach to Model Ductile Brittle Transition in Reactor Pressure Vessel Steels

    SciTech Connect

    Pritam Chakraborty; S. Bulent Biner

    2014-08-01

    In this study, a unified cohesive zone model has been proposed to predict, Ductile to Brittle Transition, DBT, in Reactor Pressure Vessel, RPV, steels. A general procedure is described to obtain the Cohesive Zone Model, CZM, parameters for the different temperatures and fracture probabilities. In order to establish the full master-curve, the procedure requires three calibration points with one at the upper-shelf for ductile fracture and two for the fracture probabilities, Pf, of 5% and 95% at the lower-shelf. In the current study, these calibrations were carried out by utilizing the experimental fracture toughness values and flow curves. After the calibration procedure, the simulations of fracture behavior (ranging from completely unstable to stable crack extension behavior) in one inch thick compact tension specimens at different temperatures yielded values that were comparable to the experimental fracture toughness values, indicating the viability of such unified modeling approach.

  10. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  11. Marine transportation and burial of the Shippingport reactor pressure vessel/neutron shield tank package

    SciTech Connect

    Coughlin, P.J.

    1989-01-01

    The Shippingport Station Decommissioning Project (SSDP) is a US Department of Energy (DOE) project for dismantling the Shippingport atomic power station. One of the more significant and challenging technical aspects of the project, which is being managed for DOE by General Electric-Nuclear Energy, is the marine transport of the reactor pressure vessel (RPV) and its associated neutron shield tank (NST) to the government-owned Hanford Reservation near Richland, Washington. Planning of the transport activity, barge transportation operations, and Hanford transportation operations, are discussed. This work will be the first use of barge transportation in the United States of a radioactive RPV package from a decommissioned land-based nuclear power plant. This extensive transportation operation has been accomplished in a timely, safe, and cost-effective manner.

  12. Prediction of the effects of thermal ageing on the embrittlement of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Margolin, B. Z.; Yurchenko, E. V.; Morozov, A. M.; Chistyakov, D. A.

    2014-04-01

    A new method has been proposed for prediction of the effects of thermal ageing on the embrittlement of reactor pressure vessel (RPV) steels. The method is based on the test results for materials in two conditions, namely, aged at temperatures of temper embrittlement and annealed after irradiation. The prediction is based on the McLean's equation and the dependencies describing thermally activated and radiation-enhanced phosphorus diffusion. Experimental studies have been carried out for estimation of thermal ageing of the WWER-1000 RPV 2Cr-Ni-Mo-V steel. The ductile to brittle transition temperature shift ΔTk due to phosphorus segregation has been estimated on the basis of experimental data processed by the proposed method for the time t = 5 × 105 h (more than 60 years of operation) for the base and weld metals of the WWER-1000 RPV.

  13. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads.

  14. Predictive Reactor Pressure Vessel Steel Irradiation Embrittlement Models: Issues and Opportunities

    SciTech Connect

    Odette, George Robert; Nanstad, Randy K

    2009-01-01

    Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature ( T) of reactor pressure vessel (RPV) steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated T correlation models that provide excellent statistical fi ts to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low fl ux-high fl uence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of late blooming phases in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict T attenuation away from the reactor core, verifi cation of the master curve method to directly measure the fracture toughness with small specimens and predicting T for vessel annealing remediation and re-irradiation cycles.

  15. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

  16. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  17. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGESBeta

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  18. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    SciTech Connect

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  19. Photofission Analysis for Fissile Dosimeters Dedicated to Reactor Pressure Vessel Surveillance

    NASA Astrophysics Data System (ADS)

    Bourganel, Stéphane; Faucher, Margaux; Thiollay, Nicolas

    2016-02-01

    Fissile dosimeters are commonly used in reactor pressure vessel surveillance programs. In this paper, the photofission contribution is analyzed for in-vessel 237Np and 238U fissile dosimeters in French PWR. The aim is to reassess this contribution using recent tools (the TRIPOLI-4 Monte Carlo code) and latest nuclear data (JEFF3.1.1 and ENDF/B-VII nuclear libraries). To be as exhaustive as possible, this study is carried out for different configurations of fissile dosimeters, irradiated inside different kinds of PWR: 900 MWe, 1300 MWe, and 1450 MWe. Calculation of photofission rate in dosimeters does not present a major problem using the TRIPOLI-4® Monte Carlo code and the coupled neutron-photon simulation mode. However, preliminary studies were necessary to identify the origin of photons responsible of photofissions in dosimeters in relation to the photofission threshold reaction (around 5 MeV). It appears that the main contribution of high enough energy photons generating photofissions is the neutron inelastic scattering in stainless steel reactor structures. By contrast, 137Cs activity calculation is not an easy task since photofission yield data are known with high uncertainty.

  20. Rate Coefficient Determinations for H + NO2 → OH + NO from High Pressure Flow Reactor Measurements.

    PubMed

    Haas, Francis M; Dryer, Frederick L

    2015-07-16

    Rate coefficients for the reaction H + NO2 → OH + NO (R1) have been determined over the nominal temperature and pressure ranges of 737-882 K and 10-20 atm, respectively, from measurements in two different flow reactor facilities: one laminar and one turbulent. Considering the existing database of experimental k1 measurements, the present conditions add measurements of k1 at previously unconsidered temperatures between ∼820-880 K, as well as at pressures that exceed existing measurements by over an order of magnitude. Experimental measurements of NOx-perturbed H2 oxidation have been interpreted by a quasi-steady state NOx plateau (QSSP) method. At the QSSP conditions considered here, overall reactivity is sensitive only to the rates of R1 and H + O2 + M → HO2 + M (R2.M). Consequently, the ratio of k1 to k2.M may be extracted as a simple algebraic function of measured NO2, O2, and total gas concentrations with only minimal complication (within measurement uncertainty) due to treatment of overall gas composition M that differs slightly from pure bath gas B. Absolute values of k1 have been determined with reference to the relatively well-known, pressure-dependent rate coefficients of R2.B for B = Ar and N2. Rate coefficients for the title reaction determined from present experimental interpretation of both laminar and turbulent flow reactor results appear to be in very good agreement around a representative value of 1.05 × 10(14) cm(3) mol(-1) s(-1) (1.74 × 10(-10) cm(3) molecule(-1) s(-1)). Further, the results of this study agree both with existing low pressure flash photolysis k1 determinations of Ko and Fontijn (J. Phys. Chem. 95 3984) near 760 K as well as a present fit to the theoretical expression of Su et al. (J. Phys. Chem. A 106 8261). These results indicate that, over the temperature range considered in this study and up to at least 20 atm, net chemistry due to stabilization of the H-NO2 reaction intermediate to form isomers of HNO2 may proceed at

  1. A Teacher's Notebook: Russian.

    ERIC Educational Resources Information Center

    National Association of Independent Schools, Boston, MA.

    Articles written by teachers of Russian for their peers in similar classroom environments are intended to aid in the development of curricular innovation. The articles, all based on practical and direct experience, include: (1) "Direct Methods for Teaching Russian," (2) "The Eclectic Approach to Teaching Russian," (3) "So-Called Linguistic…

  2. Self-consistent microwave field and plasma discharge simulations for a moderate pressure hydrogen discharge reactor

    NASA Astrophysics Data System (ADS)

    Hassouni, K.; Grotjohn, T. A.; Gicquel, A.

    1999-07-01

    A self-consistent two-dimensional model of the electromagnetic field and the plasma in a hydrogen discharge system has been developed and tested in comparison to experimental measurements. The reactor studied is a 25 cm diameter resonant cavity structure operating at 2.45 GHz with a silica belljar of 10 cm diameter and 17 cm height contained within the microwave cavity. The inside of the belljar where the discharge occurs contains a substrate holder of 5 cm diameter that is used to hold substrates for diamond deposition. The electromagnetic field model solves for the microwave fields using a finite difference time-domain solution of Maxwell's equations. The plasma model is a three energy mode (gas, molecular vibration, and electron) and nine species (H2, H, H(n=2), H(n=3), H+, H2+, H3+, H-, electron) model which accounts for non-Boltzmann electron distribution function and has 35 reactions. Simulated characteristics of the reactor in two dimensions include gas temperature, electron temperature, electron density, atomic hydrogen molar fraction, microwave power absorption, and microwave fields. Comparisons of the model are made with close agreement to several experimental measurements including coherent anti-Stokes Raman Spectroscopy measurement of H2 temperature versus position above the substrate, Doppler broadening optical emission spectroscopy (OES) measurements of H temperature versus pressure, actinometry measurements of the relative H atom concentration, Hα OES intensity measurements versus position, and microwave electric field measurements. The parameter range studied includes pressures of 2500-11 000 Pa, microwave powers of 300-2000 W, and three vertical positions of the substrate holder.

  3. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    SciTech Connect

    Boucau, Joseph; Mirabella, C.; Nilsson, Lennart; Kreitman, Paul J.; Obert, Estelle

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  4. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 1-Summary

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized- water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SAS2H analytical sequence in SCALE-4. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code sequence was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE-4 criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for analysis of each critical configuration. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the

  5. Aging of the containment pressure boundary in light-water reactor plants

    SciTech Connect

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized.

  6. 3D magnetohydrodynamic modelling of a dc low-current plasma arc batch reactor at very high pressure in helium

    NASA Astrophysics Data System (ADS)

    Lebouvier, A.; Iwarere, S. A.; Ramjugernath, D.; Fulcheri, L.

    2013-04-01

    This paper deals with a three-dimensional (3D) time-dependent magnetohydrodynamic (MHD) model under peculiar conditions of very high pressures (from 2 MPa up to 10 MPa) and low currents (<1 A). Studies on plasma arc working under these unusual conditions remain almost unexplored because of the technical and technological challenges to develop a reactor able to sustain a plasma at very high pressures. The combined effect of plasma reactivity and high pressure would probably open the way towards new promising applications in various fields: chemistry, lightning, materials or nanomaterial synthesis. A MHD model helps one to understand the complex and coupled phenomena surrounding the plasma which cannot be understood by simply experimentation. The model also provides data which are difficult to directly determine experimentally. The model simulates an experimental-based batch reactor working with helium. The particular reactor in question was used to investigate the Fischer-Tropsch application, fluorocarbon production and CO2 retro-conversion. However, as a first approach in terms of MHD, the model considers the case for helium as a non-reactive working gas. After a detailed presentation of the model, a reference case has been fully analysed (P = 8 MPa, I = 0.35 A) in terms of physical properties. The results show a bending of the arc and displacement of the anodic arc root towards the top of the reactor, due to the combined effects of convection, gravity and electromagnetic forces. A parametric study on the pressure (2-10 MPa) and current (0.25-0.4 A) was then investigated. The operating pressure does not show an influence on the contraction of the arc but higher pressures involve a higher natural convection in the reactor, driven by the density gradients between the cold and hot gas.

  7. Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

    NASA Astrophysics Data System (ADS)

    Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

    2014-12-01

    Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

  8. Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

    SciTech Connect

    Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

    1995-12-31

    Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system.

  9. Fabrication Flaw Density and Distribution in the Repairs of Reactor Pressure Vessels

    SciTech Connect

    Schuster, George J.; Doctor, Steven R.; Simonen, Fredric A.

    2006-02-15

    The Pacific Northwest National Laboratory (PNNL) is developing a generalized flaw size and density distribution for the population of U.S. reactor pressure vessels (RPVs). The purpose of the generalized flaw distribution is to predict vessel specific flaw rates for use in probabilistic fracture mechanics calculations that estimate vessel failure probability. Considerable progress has been made on the construction of an engineering data base of fabrication flaws in U.S. nuclear RPVs. The fabrication processes and product forms used to construct U.S. RPVs are represented in the data base. A validation methodology has been developed for characterizing the flaws for size, shape, orientation, and composition. The relevance of construction records has been established for describing fabrication processes and product forms. The fabrication flaws were detected in material removed from cancelled nuclear power plants using high sensitivity nondestructive ultrasonic testing, and validated by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing. This paper describes research that has generated data on welding flaws, which indicated that the largest flaws occur in weld repairs. Recent research results confirm that repair flaws are complex in composition and may include cracks on the repair ends. Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for nuclear power plant components requires radiographic examinations (RT) of welds and requires repairs for RT indications that exceed code acceptable sizes. PNNL has previously obtained the complete construction records for two RPVs. Analysis of these records show a significant change in repair frequency.

  10. Simulation study of nanoparticle coating in a low pressure plasma reactor

    SciTech Connect

    Pourali, N.; Foroutan, G.

    2015-02-15

    A self-consistent combination of plasma fluid model, nanoparticle heating model, and surface deposition model is used to investigate the coating of nanosize particles by amorphous carbon layers in a low pressure plasma reactor. The numerical results show that, owing to the net heat release in the surface reactions, the particle temperature increases and its equilibrium value remains always 50 K above the background gas temperature. The deposition rate decreases with increasing of the particle temperature and the corresponding time scale is of the order of 10 ms. The deposition rate is also strongly affected by the change in plasma parameters. When the electron temperature is increased, the deposition rate first increases due to the enhanced ion and radical generation, shows a maximum and then declines as the particle temperature rises above the gas temperature. An enhancement in the background gas pressure and/or temperature leads to a reduction in the deposition rate, which can be explained in terms of the enhanced etching by atomic hydrogen and particle heating by the background gas.

  11. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    SciTech Connect

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y.; Rashid, Y.R.

    1998-08-01

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  12. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    NASA Technical Reports Server (NTRS)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  13. Effects of gap and elevated pressure on ethanol reforming in a non-thermal plasma reactor

    NASA Astrophysics Data System (ADS)

    Hoang, Trung Q.; Zhu, Xinli; Lobban, Lance L.; Mallinson, Richard G.

    2011-07-01

    Production of hydrogen for fuel cell vehicles, mobile power generators and for hydrogen-enhanced combustion from ethanol is demonstrated using energy-efficient non-thermal plasma reforming. A tubular reactor with a multipoint electrode system operated in pulsed mode was used. Complete conversion can be achieved with high selectivity (based on ethanol) of H2 and CO of 111% and 78%, respectively, at atmospheric pressure. An elevated pressure of 15 psig shows improvement of selectivity of H2 and CO to 120% and 87%, with a significant reduction of C2Hx side products. H2 selectivity increased to 127% when a high ratio (29.2) of water-to-ethanol feed was used. Increasing CO2 selectivity is observed at higher water-to-ethanol ratios indicating that the water gas shift reaction occurs. A higher productivity and lower C2Hx products were observed at larger gas gaps. The highest overall energy efficiency achieved, including electrical power consumption, was 82% for all products or 66% for H2 only.

  14. Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

    SciTech Connect

    Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. )

    1992-03-01

    The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the primary acceptance criterion'' in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

  15. Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

    SciTech Connect

    Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K.

    1992-03-01

    The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the ``primary acceptance criterion`` in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

  16. High temperature postirradiation materials performance of spent pressurized water reactor fuel rods under dry storage conditions

    SciTech Connect

    Einziger, R.E.; Atkin, S.D.; Pasupathi, V.; Stellrecht, D.E.

    1982-04-01

    Postirradiation studies on failure mechanisms of well-characterized pressurized water reactor rods were conducted for up to a year at 482, 510, and 571/sup 0/C in limited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding, which decreases the internal rod pressure. The cladding creep also contributes to radial cracks, through the external oxide and internal fuel-cladding chemical interaction layers, which propagated into and arrested in an oxygen stabilized alpha-Zircaloy layer. There were no signs of either additional cladding hydriding, stress corrosion cracking, or fuel pellet degradation. If irradiation hardening does not reduce the stress rupture properties of Zircaloy, a conservative maximum storage temperature of 400/sup 0/C based on a stress-rupture mechanism is recommended to ensure a 1000-yr cladding lifetime.

  17. Ultimate pressure capacity analysis of a post-tensioned reinforced concrete nuclear reactor containment building

    SciTech Connect

    Dooley, W.T.; Macek, R.W.; Sadik, S.

    1983-01-01

    This paper presents a three dimensional nonlinear structural analysis of a post-tensioned reinforced concrete nuclear reactor containment building. The objective of the analysis was to develop and demonstrate modeling techniques appropriate for determining the global ultimate internal pressure capacity of this type of containment. The structural model developed for the ADINA computer code employed nonlinear material models with truss, shell and three dimensional continuum elements to represent the major structural members (tendons, rebars, ties, concrete and liner). In addition a special uniaxial elastic plastic shell material model was developed to facilitate representation of the dome tendon lacing pattern. Effective use of symmetry permitted all three dimensional effects to be represented in a 30/sup 0/ (0.52 rad) segment of the structure. The analysis suggested that gross failure of the structure at a quasistatic internal pressure of 99 psi (0.68 MPa) may have been precipitated by the nonaxisymmetric three dimensional nature of the stress field in the concrete near the tendon anchorages.

  18. Investigation of the DSMC Approach for Ion/neutral Species in Modeling Low Pressure Plasma Reactor

    SciTech Connect

    Deng Hao; Li, Z.; Levin, D.; Gochberg, L.

    2011-05-20

    Low pressure plasma reactors are important tools for ionized metal physical vapor deposition (IMPVD), a semiconductor plasma processing technology that is increasingly being applied to deposit Cu seed layers on semiconductor surfaces of trenches and vias with the high aspect ratio (e.g., >5:1). A large fraction of ionized atoms produced by the IMPVD process leads to an anisotropic deposition flux towards the substrate, a feature which is critical for attaining a void-free and uniform fill. Modeling such devices is challenging due to their high plasma density, reactive environment, but low gas pressure. A modular code developed by the Computational Optical and Discharge Physics Group, the Hybrid Plasma Equipment Model (HPEM), has been successfully applied to the numerical investigations of IMPVD by modeling a hollow cathode magnetron (HCM) device. However, as the development of semiconductor devices progresses towards the lower pressure regime (e.g., <5 mTorr), the breakdown of the continuum assumption limits the application of the fluid model in HPEM and suggests the incorporation of the kinetic method, such as the direct simulation Monte Carlo (DSMC), in the plasma simulation.The DSMC method, which solves the Boltzmann equation of transport, has been successfully applied in modeling micro-fluidic flows in MEMS devices with low Reynolds numbers, a feature shared with the HCM. Modeling of the basic physical and chemical processes for ion/neutral species in plasma have been developed and implemented in DSMC, which include ion particle motion due to the Lorentz force, electron impact reactions, charge exchange reactions, and charge recombination at the surface. The heating of neutrals due to collisions with ions and the heating of ions due to the electrostatic field will be shown to be captured by the DSMC simulations. In this work, DSMC calculations were coupled with the modules from HPEM so that the plasma can be self-consistently solved. Differences in the Ar

  19. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    SciTech Connect

    Griggs, D.P.

    1993-03-01

    The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification.

  20. Mechanical properties and microstructure of three Russian ferritic/martensitic steels irradiated in BN-350 reactor to 50 dpa at 490C

    SciTech Connect

    Dvoriashin, Alexander M.; Porollo, S. I.; Konobeev, Yu V.; Budylkin, N. I.; Mironova, E. G.; Ioltukhovsky, A. G.; Leonteva-Smirnova, M. V.; Garner, Francis A.

    2007-08-01

    Ferritic/martensitic (F/M) steels are being considered for application in fusion reactors, intense neutron sources, and accelerator-driven systems. While EP-450 is traditionally used with sodium coolants in Russia, EP-823 and EI-852 steels with higher silicon levels have been developed for reactor facilities using lead-bismuth coolant. To determine the influence of silicon additions on short-term mechanical properties and microstructure, ring specimens cut from cladding tubes of these three steels were irradiated in sodium at 490С in the BN-350 reactor to 50 dpa. Post-irradiation tensile testing and microstructural examination show that EI-852 steel (1.9 wt% Si) undergoes severe irradiation embrittlement. Microstructural investigation showed that the formation of near-continuous -phase precipitates on grain boundaries is the main cause of the embrittlement.

  1. Models for embrittlement recovery due to annealing of reactor pressure vessel steels

    SciTech Connect

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1995-05-01

    The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  2. Evolution of Nickel-Manganese-Silicon Dominated Phases in Highly Irradiated Reactor Pressure Vessel Steels

    SciTech Connect

    Peter B Wells; Yuan Wu; Tim Milot; G. Robert Odette; Takuya Yamamoto; Brandon Miller; James Cole

    2014-11-01

    Formation of a high density of Ni-Mn-Si nm-scale precipitates in irradiated reactor pressure vessel steels, both with and without Cu, could lead to severe embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement regulations, would emerge only at high fluence. However, the mechanisms and variables that control Ni-Mn- Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni were carried out at ˜ 295±5°C to high and very high neutron fluences of ˜ 1.3x1020 and 1.1x1021 n/cm2. Atom probe tomography (APT) shows that significant mole fractions of these precipitates form in the Cu bearing steels at ˜ 1.3x1020 n/cm2, while they are only beginning to develop in Cu-free steels. However, large mole fractions, far in excess of those found in previous studies, are observed at 1.1x1021 n/cm2 at all Cu levels. The precipitates diffract, and in one case are compositionally and structurally consistent with the Mn6Ni16Si7 G-phase. At the highest fluence, the large precipitate mole fractions primarily depend on the steel Ni content, rather than Cu, and lead to enormous strength increases up to about 700 MPa. The implications of these results to light water reactor life extension are discussed briefly.

  3. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  4. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  5. Approaches to Accounting and Prediction of Fast Neutron Fluence on VVER Pressure Vessels for Estimation of RPV Residual Lifetime in Compliance with Russian Utility's Procedure

    NASA Astrophysics Data System (ADS)

    Borodkin, Gennady; Borodkin, Pavel; Khrennikov, Nikolay; Ryabinin, Yuriy; Adeev, Valeriy

    2016-02-01

    The Paper describes a new Russian Utility's regulatory document (RD EO) which has been recently developed and implemented since the beginning of 2013. This RD EO includes the procedure of RPV FNF monitoring and provides recommendations on how to predict fluence over the design lifetime taking into account results of FNF monitoring. The basic method of RPV neutron fluence monitoring is neutron transport calculations of FR in the vicinity of the RPV. Reliability of the calculation results should be validated by ex-vessel neutron-activation measurements, which were performed during different fuel cycles with different core loadings including new types of fuel.

  6. Pressurized thermal shock: TEMPEST computer code simulation of thermal mixing in the cold leg and downcomer of a pressurized water reactor. [Creare 61 and 64

    SciTech Connect

    Eyler, L.L.; Trent, D.S.

    1984-04-01

    The TEMPEST computer program was used to simulate fluid and thermal mixing in the cold leg and downcomer of a pressurized water reactor under emergency core cooling high-pressure injection (HPI), which is of concern to the pressurized thermal shock (PTS) problem. Application of the code was made in performing an analysis simulation of a full-scale Westinghouse three-loop plant design cold leg and downcomer. Verification/assessment of the code was performed and analysis procedures developed using data from Creare 1/5-scale experimental tests. Results of three simulations are presented. The first is a no-loop-flow case with high-velocity, low-negative-buoyancy HPI in a 1/5-scale model of a cold leg and downcomer. The second is a no-loop-flow case with low-velocity, high-negative density (modeled with salt water) injection in a 1/5-scale model. Comparison of TEMPEST code predictions with experimental data for these two cases show good agreement. The third simulation is a three-dimensional model of one loop of a full size Westinghouse three-loop plant design. Included in this latter simulation are loop components extending from the steam generator to the reactor vessel and a one-third sector of the vessel downcomer and lower plenum. No data were available for this case. For the Westinghouse plant simulation, thermally coupled conduction heat transfer in structural materials is included. The cold leg pipe and fluid mixing volumes of the primary pump, the stillwell, and the riser to the steam generator are included in the model. In the reactor vessel, the thermal shield, pressure vessel cladding, and pressure vessel wall are thermally coupled to the fluid and thermal mixing in the downcomer. The inlet plenum mixing volume is included in the model. A 10-min (real time) transient beginning at the initiation of HPI is computed to determine temperatures at the beltline of the pressure vessel wall.

  7. Core damage severity evaluation for pressurized water reactors by artificial intelligence methods

    NASA Astrophysics Data System (ADS)

    Mironidis, Anastasios Pantelis

    1998-12-01

    During the course of nuclear power evolution, accidents have occurred. However, in the western world, none of them had a severe impact on the public because of the design features of nuclear plants. In nuclear reactors, barriers constitute physical obstacles to uncontrolled fission product releases. These barriers are an important factor in safety analysis. During an accident, reactor safety systems become actuated to prevent the barriers from been breached. In addition, operators are required to take specified actions, meticulously depicted in emergency response procedures. In an accident, on-the-spot knowledge regarding the condition of the core is necessary. In order to make the right decisions toward mitigating the accident severity and its consequences, we need to know the status of the core [1, 3]. However, power plant instrumentation that can provide a direct indication of the status of the core during the time when core damage is a potential outcome, does not exist. Moreover, the information from instruments may have large uncertainty of various types. Thus, a very strong potential for misinterpreting incoming information exists. This research endeavor addresses the problem of evaluating the core damage severity of a Pressurized Water Reactor during a transient or an accident. An expert system has been constructed, that incorporates knowledge and reasoning of human experts. The expert system's inference engine receives incoming plant data that originate in the plethora of core-related instruments. Its knowledge base relies on several massive, multivariate fuzzy logic rule-sets, coupled with several artificial neural networks. These mathematical models have encoded information that defines possible core states, based on correlations of parameter values. The inference process classifies the core as intact, or as experiencing clad damage and/or core melting. If the system detects a form of core damage, a quantification procedure will provide a numerical

  8. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... reactors where neutron radiation has reduced the fracture toughness of the reactor vessel materials, a thermal annealing may be applied to the reactor vessel to recover the fracture toughness of the material....4 at least three years prior to the date at which the limiting fracture toughness criteria in §...

  9. Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven

    SciTech Connect

    Spencer, Benjamin; Hoffman, William; Sen, Sonat; Rabiti, Cristian; Dickson, Terry; Bass, Richard

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  10. Development of Alternative Reductant Application in Pressurized Water Reactor Primary Systems

    NASA Astrophysics Data System (ADS)

    Domae, Masafumi; Kawamura, Hirotaka; Ohira, Taku

    In primary coolant of pressurized water reactors, high concentration dissolved hydrogen (DH) has been added, to prevent generation of oxidizing species through radiolysis of the coolant. Recently, number of ageing plants is increasing and utilities are concerned about primary water stress corrosion cracking (PWSCC). Some researchers consider that occurrence of PWSCC and crack propagation rate are affected by the DH concentration. The authors consider that one of possible mitigation methods toward PWSCC is use of alternative reductant in place of hydrogen. Because from the radiation chemical aspect aliphatic alcohols are typical scavengers of the oxidizing radical generated through the radiolysis of water, they are promising candidates of the alternative reductant. In the present work, possible alternatives of hydrogen were screened, and methanol was selected as the best candidate. Corrosion tests of type 304 stainless steels were carried out in high temperature water at 320°C without irradiation under two conditions: (1) DH 1.5 ppm (part per million) and (2) methanol 2.9 ppm. Electrochemical corrosion potential of the stainless steel specimens was measured during the immersion tests. After the immersion tests for 1500 h, surface morphology of the stainless steel specimens was observed by scanning probe microscope. Major component of the oxide film formed on the stainless steel specimens was analyzed by X-ray diffraction. From comparison of the test results, it is concluded that addition of 2.9 ppm methanol has almost the same effect on corrosion environment as DH 1.5 ppm addition.

  11. Crack arrest behavior of reactor pressure vessel steels at high temperatures

    SciTech Connect

    Pugh, C.E.; Naus, D.J.; Bass, B.R.

    1988-01-01

    The Heavy-Section Steel Technology Program at the Oak Ridge National Laboratory under the sponsorship of the US Nuclear Regulatory Commission is conducting experimental and analytical studies to improve the understanding of conditions that govern the initiation, rapid propagation, arrest and ductile tearing of cracks in reactor pressure vessel (RPV) steels. In support of this objective, large-scale wide-plate experiments are performed to generate crack-arrest toughness data for RPV steels at temperatures approaching and above the onset of Charpy upper-shelf behavior. Analytical studies are addressing the role of dynamics and nonlinear rate-dependent (i.e., viscoplastic) effects in the interpretation of crack run-arrest events in these ductile materials. A summary of the wide-plate tests performed to date is presented, including details of test procedures, test data, and results of analyses performed to date. The importance of incorporating viscoplastic effects into dynamic analysis of crack run-arrest events in these strain-rate sensitive steels is examined through applications of selected proposed viscoplastic constitutive equations and fracture parameters to the interpretation of data from the wide-plate tests. The crack-arrest data are compared with those from small ASTM-type specimens and other large structural tests.

  12. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect

    Server, W. L.; Nanstad, Randy K

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  13. Iron catalyst chemistry in modeling a high-pressure carbon monoxide nanotube reactor

    NASA Technical Reports Server (NTRS)

    Scott, Carl D.; Povitsky, Alexander; Dateo, Christopher; Gokcen, Tahir; Willis, Peter A.; Smalley, Richard E.

    2003-01-01

    The high-pressure carbon monoxide (HiPco) technique for producing single-wall carbon nanotubes (SWNTs) is analyzed with the use of a chemical reaction model coupled with flow properties calculated along streamlines, calculated by the FLUENT code for pure carbon monoxide. Cold iron pentacarbonyl, diluted in CO at about 30 atmospheres, is injected into a conical mixing zone, where hot CO is also introduced via three jets at 30 degrees with respect to the axis. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Then iron nucleates and forms clusters that catalyze the formation of SWNTs by a disproportionation reaction (Boudouard) of CO on Fe-containing clusters. Alternative nucleation rates are estimated from the theory of hard sphere collision dynamics with an activation energy barrier. The rate coefficient for carbon nanotube growth is estimated from activation energies in the literature. The calculated growth was found be about an order of magnitude greater than measured, regardless of the nucleation rate. A study of cluster formation in an incubation zone prior to injection into the reactor shows that direct dimer formation from Fe atoms is not as important as formation via an exchange reaction of Fe with CO in FeCO.

  14. Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA GRC

    NASA Technical Reports Server (NTRS)

    Reid, Terry V.

    2015-01-01

    Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA GRC. The TDU consists of three subsystems: the Reactor Simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the Heat Exchanger Manifold (HXM). An Annular Linear Induction Pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now), is referred to as the RxSim subsystem and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the CAD model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras turbulence model), and 2.223 kg/sec (using the k-? turbulence model). The computational error of the predictions for the available mass flow is -0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k-epsilon turbulence model) when compared to measured data.

  15. Synthesis of amorphous silicon carbide nanoparticles in a low temperature low pressure plasma reactor.

    PubMed

    Lin, Hongfei; Gerbec, Jeffrey A; Sushchikh, Michael; McFarland, Eric W

    2008-08-13

    Commercial scale production of silicon carbide (SiC) nanoparticles smaller than 10 nm remains a significant challenge. In this paper, a microwave plasma reactor and appropriate reaction conditions have been developed for the synthesis of amorphous SiC nanoparticles. This continuous gas phase process is amenable to large scale production use and utilizes the decomposition of tetramethylsilane (TMS) for both the silicon and the carbon source. The influence of synthesis parameters on the product characteristics was investigated. The as-prepared SiC particles with sizes between 4 and 6 nm were obtained from the TMS precursor in a plasma operated at low temperature and low precursor partial pressure (0.001-0.02 Torr) using argon as the carrier gas (3 Torr). The carbon:silicon ratio was tuned by the addition of hydrogen and characterized by x-ray photoelectron spectroscopy. The reaction mechanism of SiC nanoparticle formation in the microwave plasma was investigated by mass spectroscopy of the gaseous products. PMID:21828814

  16. Warm PreStress effect on highly irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Hure, J.; Vaille, C.; Wident, P.; Moinereau, D.; Landron, C.; Chapuliot, S.; Benhamou, C.; Tanguy, B.

    2015-09-01

    This study investigates the Warm Prestress (WPS) effect on 16MND5 (A508 Cl3) RPV steel, irradiated up to a fluence of 13 ·1023 n .m-2 (E > 1 MeV) at a temperature of 288 ° C, corresponding to more than 60 years of operations in a French Pressurized Water Reactor (PWR). Mechanical properties, including tensile tests with different strain rates and tension-compression tests on notched specimens, have been characterized at unirradiated and irradiated states and used to calibrate constitutive equations to describe the mechanical behavior as a function of temperature and fluence. Irradiation embrittlement has been determined based on Charpy V-notch impact tests and isothermal quasi-static toughness tests. Assessment of WPS effect has been done through various types of thermomechanical loadings performed on CT(0.5 T) specimens. All tests have confirmed the non-failure during the thermo-mechanical transients. Experimental data obtained in this study have been compared to both engineering-based models and to a local approach (Beremin) model for cleavage fracture. It is shown that both types of modeling give good predictions for the effective toughness after warm prestressing.

  17. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    SciTech Connect

    McHenry, H.I.; Alers, G.A.

    1998-03-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

  18. Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

    NASA Astrophysics Data System (ADS)

    Niffenegger, M.; Leber, H. J.

    2009-06-01

    The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient ( K¯) of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10 19 neutrons per cm 2, are presented. Within this range, it was observed that K¯ increased by ≈500 nV/°C and a linear dependency was noted between K¯ and the temperature shift Δ T41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.

  19. Decision process involved in preparing the Shippingport reactor pressure vessel for transport

    SciTech Connect

    Murphie, W.E.

    1989-01-01

    The most significant part of the Shippingport Station Decommissioning Project was the one-piece removal and shipment of the reactor pressure vessel (RPV). Implicit in the RPV transport was the task of qualifying the RPV as a waste package acceptable for shipment. Soon after physical decommissioning began on September 1985, questions regarding the packaging certification and transport of the RPV from Shippingport, Pennsylvania to the US Department of Energy (DOE) Hanford Waste Burial Site necessitated reexamination of several planning assumptions. A complete reassessment of the regulatory requirements governing the RPV shipment resulted in a programmatic decision to obtain a type B(U) Certificate of Compliance and abandon the originally planned US Department of Transportation (DOT) low specific activity (LSA) shipment. The decision process resulting in this conclusion was extensive and involved many organizations and agencies. Incidental to this process, several subtle certification issues were identified that required resolution. Some of these issues involved the definition of LSA material for large packages; interpretation and compliance with DOE, DOT and US Nuclear Regulatory Commission (NRC) regulations for the transport of radioactive material; incorporation of the International Atomic Energy Agency (IAEA) regulations by the Panama Canal; and DOE policy requiring advance notification to states of radioactive waste shipments. 2 figs.

  20. Generic risk insights for Westinghouse and Combustion Engineering pressurized water reactors

    SciTech Connect

    Travis, R.; Taylor, J.; Fresco, A. ); Chung, J. )

    1990-11-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of Westinghouse and Combustion Engineering (CE) pressurized water reactors (PWRs) and apply the insights gained to Westinghouse and Ce plants have not been subjected to a PRA. The available PRAs (five Westinghouse plants and one CE plant) were examined to identify the most probable, i.e., dominant accident sequences at each plant. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eleven sequences met this definition. From these sequences, the most important component failures and human errors that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training maintenance, design review, and inspections.

  1. Implosion of reactor-size, gas-filled spherical shell targets driven by shaped pressure pulses

    SciTech Connect

    Piriz, A.R.; Atzeni, S. )

    1993-05-01

    The implosion of a family of reactor-size targets for inertial confinement fusion (ICF) is studied analytically and numerically. The targets consist of a deuterium--tritium (D--T) shell filled with D--T vapor and they are imploded by a multistep pressure pulse designed in such a way that the final hot spot is formed mainly from the initially gaseous fuel. The formation of the hot spot is described by means of a relatively simple model, and scaling laws for the quantities that characterize the state of the initially gaseous part of the fuel prior to ignition are derived. The results of the model are compared with one-dimensional fluid simulations, and good agreement is found. A parametric study of the fuel energy gain is then presented; the dependence of the gain and of the hot spot convergence ratio on the pulse parameters and on the filling gas density is analyzed. It is also shown that a substantial increase in the gain (for a given target and pulse energy) can be achieved by replacing the last step of the pulse with an exponential ramp.

  2. Iron catalyst chemistry in modeling a high-pressure carbon monoxide nanotube reactor.

    PubMed

    Scott, Carl D; Povitsky, Alexander; Dateo, Christopher; Gökçen, Tahir; Willis, Peter A; Smalley, Richard E

    2003-01-01

    The high-pressure carbon monoxide (HiPco) technique for producing single-wall carbon nanotubes (SWNTs) is analyzed with the use of a chemical reaction model coupled with flow properties calculated along streamlines, calculated by the FLUENT code for pure carbon monoxide. Cold iron pentacarbonyl, diluted in CO at about 30 atmospheres, is injected into a conical mixing zone, where hot CO is also introduced via three jets at 30 degrees with respect to the axis. Hot CO decomposes the Fe(CO)5 to release atomic Fe. Then iron nucleates and forms clusters that catalyze the formation of SWNTs by a disproportionation reaction (Boudouard) of CO on Fe-containing clusters. Alternative nucleation rates are estimated from the theory of hard sphere collision dynamics with an activation energy barrier. The rate coefficient for carbon nanotube growth is estimated from activation energies in the literature. The calculated growth was found be about an order of magnitude greater than measured, regardless of the nucleation rate. A study of cluster formation in an incubation zone prior to injection into the reactor shows that direct dimer formation from Fe atoms is not as important as formation via an exchange reaction of Fe with CO in FeCO. PMID:12908231

  3. Development of Pressurized Water Reactor Integrated Safety Analysis Methodology Using Multilevel Coupling Algorithm

    SciTech Connect

    Ziabletsev, Dmitri; Avramova, Maria; Ivanov, Kostadin

    2004-11-15

    The subchannel code COBRA-TF has been introduced for an evaluation of thermal margins on the local pin-by-pin level in a pressurized water reactor. The coupling of COBRA-TF with TRAC-PF1/NEM is performed by providing from TRAC to COBRA-TF axial and radial thermal-hydraulic boundary conditions and relative pin-power profiles, obtained with the pin power reconstruction model of the nodal expansion method (NEM). An efficient algorithm for coupling of the subchannel code COBRA-TF with TRAC-PF1/NEM in the parallel virtual machine environment was developed addressing the issues of time synchronization, data exchange, spatial overlays, and coupled convergence. Local feedback modeling on the pin level was implemented into COBRA-TF, which enabled updating the local form functions and the recalculation of the pin powers in TRAC-PF1/NEM after obtaining the local feedback parameters. The coupled TRAC-PF1/NEM/COBRA-TF code system was tested on the rod ejection accident and main steam line break benchmark problems. In both problems, the local results are closer than before the introduced multilevel coupling to the corresponding critical limits. This fact indicates that the assembly average results tend to underestimate the accident consequences in terms of local safety margins. The capability of local safety evaluation, performed simultaneously (online) with coupled global three-dimensional neutron kinetics/thermal-hydraulic calculations, is introduced and tested. The obtained results demonstrate the importance of the current work.

  4. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Li, C. W.; Han, L. Z.; Luo, X. M.; Liu, Q. D.; Gu, J. F.

    2016-08-01

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe3C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo2C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained.

  5. Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs

    SciTech Connect

    Sienicki, J.J.; Horak, W.C.; Brookhaven National Lab., Upton, NY )

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.

  6. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  7. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  8. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  9. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  10. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    DOE PAGESBeta

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-06-29

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

  11. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    SciTech Connect

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs.

  12. Observation of pressure gradient and related flow rate effect on the plasma parameters in plasma processing reactor

    SciTech Connect

    Lee, Hyo-Chang; Kim, Aram; Chung, Chin-Wook; Moon, Se Youn

    2011-02-15

    In industrial plasma processes, flow rate has been known to a key to control plasma processing results and has been discussed with reactive radical density, gas residence time, and surface reaction. In this study, it was observed that the increase in the flow rate can also change plasma parameters (electron temperature and plasma density) and electron energy distribution function in plasma processing reactor. Based on the measurement of gas pressure between the discharge region and the pumping port region, the considerable differences in the gas pressure between the two regions were found with increasing flow rate. It was also observed that even in the discharge region, the pressure gradient occurs at the high gas flow rate. This result shows that increasing the flow rate results in the pressure gradient and causes the changes in the plasma parameters.

  13. Technology, safety and costs of decommissioning a reference pressurized water reactor power station. Classification of decommissioning wastes. Addendum 3

    SciTech Connect

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference pressurized water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 17,885 cubic meters of waste from DECON are classified as follows: Class A, 98.0%; Class B, 1.2%; Class C, 0.1%. About 0.7% (133 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

  14. Effects of crack tip plastic zone on corrosion fatigue cracking of alloy 690(TT) in pressurized water reactor environments

    NASA Astrophysics Data System (ADS)

    Xiao, J.; Qiu, S. Y.; Chen, Y.; Fu, Z. H.; Lin, Z. X.; Xu, Q.

    2015-01-01

    Alloy 690(TT) is widely used for steam generator tubes in pressurized water reactor (PWR), where it is susceptible to corrosion fatigue. In this study, the corrosion fatigue behavior of Alloy 690(TT) in simulated PWR environments was investigated. The microstructure of the plastic zone near the crack tip was investigated and labyrinth structures were observed. The relationship between the crack tip plastic zone and fatigue crack growth rates and the environment factor Fen was illuminated.

  15. [Safety systems in nuclear power plants with pressurized water reactors and demonstration of their function in technical scale experiments. II].

    PubMed

    Hennies, H H

    1987-11-01

    Analyses and experiments carried out during the last decade on the sequence and consequences of accidents in German pressurized water reactors have shown that the functioning capability of the safety systems is guaranteed for the case of the MCA, the maximum credible accident. For the case of core meltdown, simulation experiments have also made it evident that the consequences remain largely restricted to the plant proper. PMID:3431585

  16. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    SciTech Connect

    Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph; Jubin, Robert Thomas

    2015-09-01

    will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. Samples of Surry-2 and H.B. Robinson pressurized water reactor cladding were heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. Cladding samples were also heated within the temperature range of 480–600ºC expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. The tritium content of the Surry-2 and H.B. Robinson cladding was measured to be ~234 and ~500 µCi/g, respectively. Heating the Surry-2 cladding at 500°C for 24 h removed ~0.2% of the tritium from the cladding, and heating at 700°C for 24 h removed ~9%. Heating the H.B. Robinson cladding at 700°C for 24 h removed ~11% of the tritium. When samples of the Surry-2 and H.B. Robinson claddings were heated at 700°C for 96 h, essentially all of the tritium in the cladding was removed. However, only ~3% of the tritium was removed when a sample of Surry-2 cladding was heated at 600°C for 96 h. These data indicate that the amount of tritium released from tritium pretreatment systems will be dependent on both the operating temperature and length of time in the system. Under certain conditions, a significant fraction of the tritium could remain bound in the cladding and would need to be considered in operations involving cladding recycle.

  17. Russian Language Analysis Project

    ERIC Educational Resources Information Center

    Serianni, Barbara; Rethwisch, Carolyn

    2011-01-01

    This paper is the result of a language analysis research project focused on the Russian Language. The study included a diverse literature review that included published materials as well as online sources in addition to an interview with a native Russian speaker residing in the United States. Areas of study include the origin and history of the…

  18. Russian Supplementary Dialogues.

    ERIC Educational Resources Information Center

    Peace Corps, Ashgabat (Turkmenistan).

    This manual is designed for the Russian language training of Peace Corps volunteers serving in Turkmenistan, and focuses on daily communication skills needed in that context. It consists of nine topical lessons, each containing several brief dialogues targeting specific language competencies, and exercises. Text is entirely in Russian, except for…

  19. RUSSIAN FOR SECONDARY SCHOOLS.

    ERIC Educational Resources Information Center

    New York State Education Dept., Albany.

    THE NEW YORK STATE SYLLABUS FOR RUSSIAN IN SECONDARY SCHOOLS FOLLOWS THE SAME FORMAT AS THOSE FOR FRENCH, GERMAN, AND SPANISH, AND FOR COMPLETE TEXT, INCLUDING GENERAL SECTIONS ON TEACHING LANGUAGES, THE READER MUST REFER TO ONE OF THOSE THREE BOOKS. THIS GUIDE DELINEATES THE AIMS, TECHNIQUES, CONTENT, AND SCOPE OF RUSSIAN INSTRUCTION FOR A 6-YEAR…

  20. Russian Teaching Contracts.

    ERIC Educational Resources Information Center

    Stevens, Betsy

    2000-01-01

    Analyzes two Russian teaching contracts, rhetorically comparing purpose and audience, culture, gender, and the role of the individual versus the state. Uses anecdotal episodes as a framework for examining Russian culture and analyzing university teaching contracts, concluding that the contracts are not only brief and factual but also reflect a…

  1. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  2. Time-Varying Characteristics Analysis and Fuzzy Controller Systematic Design Method for Pressurized Water Reactor Power Control

    SciTech Connect

    Liu Shengzhi; Zhang Naiyao; Cui Zhenhua

    2004-11-15

    In this paper a systematic design method of fuzzy control systems is applied to the pressurized water reactor's (PWR) power control. The paper includes three parts. In the first part, a simplified time-varying linear model of the PWR power system is constructed, and its inner structure and time-varying characteristics are analyzed. That provides a solid basis for study and design of the nuclear reactor power control system. In the second part, a systematic design method of fuzzy control systems is introduced and applied to control the nuclear reactor power process. The design procedures and parameters are given in detail. This systematic design method has some notable advantages. The control of a global fuzzy model can be decomposed into controlling a set of linear submodels. Each submodel controller can be independently designed by using a linear quadratic regulator approach. This systematic design method gives a sufficient and necessary condition to guarantee the stability of fuzzy control systems; thus, better control performance can be obtained due to the accurate control gains. In the third part, the control performance of the nuclear reactor fuzzy control system is examined by simulation experiments, including nuclear reactor power shutdown, start-up, and adjustment operations. The satisfactory experiment results have shown that the systematic design method for fuzzy control systems is effective and feasible.

  3. Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies

    NASA Astrophysics Data System (ADS)

    Feng Tao, Franklin; Nguyen, Luan; Zhang, Shiran

    2013-03-01

    Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ˜10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified.

  4. Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies

    NASA Astrophysics Data System (ADS)

    Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

    2013-03-01

    Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ~10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified.

  5. Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies.

    PubMed

    Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

    2013-03-01

    Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ∼10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified. PMID:23556828

  6. Dynamic Strain Aging in New Generation Cr-Mo-V Steel for Reactor Pressure Vessel Applications

    NASA Astrophysics Data System (ADS)

    Gupta, C.; Chakravartty, J. K.; Banerjee, S.

    2010-12-01

    A new generation nuclear reactor pressure vessel steel (CrMoV type) having compositional similarities with thick section 3Cr-Mo class of low alloy steels and adapted for nuclear applications was investigated for various manifestations of dynamic strain aging (DSA) using uniaxial tests. The steel investigated herein has undergone quenched and tempered treatment such that a tempered bainite microstructure with Cr-rich carbides was formed. The scope of the uniaxial experiments included tensile tests over a temperature range of 298 K to 873 K (25 °C to 600 °C) at two strain rates (10-3 and 10-4 s-1), as well as suitably designed transient strain rate change tests. The flow behavior displayed serrated flow, negative strain rate sensitivity, plateau behavior of yield, negative temperature ( T), and strain rate left( {dot{\\varepsilon }} right) dependence of flow stress over the temperature range of 523 K to 673 K (250 °C to 400 °C) and strain rate range of 5 × 10-3 s-1 to 3 × 10-6 s-1, respectively. While these trends attested to the presence of DSA, a lack of work hardening and near negligible impairment of ductility point to the fact that manifestations of embrittling features of DSA were significantly enervated in the new generation pressure vessel steel. In order to provide a mechanistic understanding of these unique combinations of manifestations of DSA in the steel, a new approach for evaluation of responsible solutes from strain rate change tests was adopted. From these experiments and calculation of activation energy by application of vacancy-based models, the solutes responsible for DSA were identified as carbon/nitrogen. The lack of embrittling features of DSA in the steel was rationalized as being due to the beneficial effects arising from the presence of dynamic recovery effects, presence of alloy carbides in the tempered bainitic structure, and formation of solute clusters, all of which hinder the possibilities for strong aging of dislocations.

  7. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  8. Biaxial loading effects on fracture toughness of reactor pressure vessel steel

    SciTech Connect

    McAfee, W.J.; Bass, B.R.; Bryson, J.W. Jr.; Pennell, W.E.

    1995-03-01

    The preliminary phases of a program to develop and evaluate fracture methodologies for assessing crack-tip constraint effects on fracture toughness of reactor pressure vessel (RPV) steels have been completed by the Heavy-Section Steel Technology (HSST) Program. Objectives were to investigate effect of biaxial loading on fracture toughness, quantify this effect through existing stress-based, dual-parameter, fracture-toughness correlations, or propose and verify alternate correlations. A cruciform beam specimen with 2-D, shallow, through-thickness flaw and a special loading fixture was designed and fabricated. Tests were performed using biaxial loading ratios of 0:1 (uniaxial), 0.6:1, and 1:1 (equi-biaxial). Critical fracture-toughness values were calculated for each test. Biaxial loading of 0.6:1 resulted in a reduction in the lower bound fracture toughness of {approximately}12% as compared to that from the uniaxial tests. The biaxial loading of 1:1 yielded two subsets of toughness values; one agreed well with the uniaxial data, while one was reduced by {approximately}43% when compared to the uniaxial data. Results were evaluated using J-Q theory and Dodds-Anderson (D-A) micromechanical scaling model. The D-A model predicted no biaxial effect, while the J-Q method gave inconclusive results. When applied to the 1:1 biaxial data, these constraint methodologies failed to predict the observed reduction in fracture toughness obtained in one experiment. A strain-based constraint methodology that considers the relationship between applied biaxial load, the plastic zone width in the crack plane, and fracture toughness was formulated and applied successfully to the data. Evaluation of this dual-parameter strain-based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition region, including the effects of out-of-plane loading on fracture toughness. This report is designated as HSST Report No. 150.

  9. Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

    NASA Technical Reports Server (NTRS)

    Reid, Terry V.

    2016-01-01

    Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S?A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ?0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

  10. Distribution of sulfur during coal pyrolysis in a high pressure entrained-flow reactor

    SciTech Connect

    Fatemi-Badi, M.; Scaroni, A.W.; Jenkins, R.G. )

    1988-01-01

    Desulfurization of coal prior to combustion can be achieved by one or more of the following methods. (1) Advanced physical cleaning--this method involves a variety of approaches for reduction of pyrite in the coal based on the differences between the density of pyrite or marcasite and the organic matrix. (2) Chemical coal cleaning--based on the concept of breaking the chemical bonds of the organic sulfur by exposing the coal usually to molten chemicals such as sodium hydroxide. (3) Conversion of the coal to low-sulfur liquid and gaseous fuels--based on liquefaction and gasification technologies. According to the literature, there are three commonly recognizable forms of sulfur in coal: sulfate, pyrite and organic sulfur. Although the distribution of various forms of sulfur is less often determined than the total sulfur, their presence can have a significant effect on coal utilization. However, the degree to which the sulfur can be reduced and/or removed from the coal is strongly dependent on the forms of sulfur present in the coal. Therefore, accurate analytical data on the forms of sulfur is a requirement for improving removal processes. Pyrolysis has been used by some researchers to study the behavior and distribution of sulfur in coal. Some investigators have also used pyrolysis in order to study the chemistry and kinetics of reactions involving sulfur-containing compounds in coal. The purpose of the current work was to determine the distribution of the total sulfur between the products of pyrolysis (tars and chars) and the various forms of sulfur in the chars from pyrolysis of a subbituminous coal in a high pressure entrained-flow reactor.

  11. CONFERENCES AND SYMPOSIA: On the 50th anniversary of the L F Vereshchagin Institute for High Pressure Physics, RAS (Scientific outreach session of the Physical Sciences Division of the Russian Academy of Sciences, 23 April 2008)

    NASA Astrophysics Data System (ADS)

    Stishov, S. M.; Khvostantsev, L. G.; Slesarev, V. N.; Popova, S. V.; Brazhkin, V. V.; Dyuzheva, T. I.; Dzhavadov, L. N.; Gromnitskaya, E. L.; Stepanov, G. N.; Timofeev, Yu A.; Dizhur, E. M.; Venttsel, V. A.; Voronovskii, A. N.; Ryzhov, V. N.; Barabanov, A. F.; Magnitskaya, M. V.; Tareeva, E. E.

    2008-10-01

    A scientific outreach session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) was held on 23 April 2008 at the Institute for High Pressure Physics, RAS, Troitsk, Moscow region. The session was devoted to the 50th anniversary of the Institute. The following reports were presented: (1) Stishov S M (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "The Institute for High Pressure Physics is now 50 (opening address)"; (2) Khvostantsev L G and Slesarev V N (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "Large-volume high-pressure devices for physical investigations"; (3) Popova S V, Brazhkin V V and Dyuzheva T I (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "Structural phase transitions in highly compressed substances and the synthesis of high-pressure phases"; (4) Dzhavadov L N, Gromnitskaya E L, Stepanov G N and Timofeev Yu A (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "Studies of the thermodynamic, elastic, superconducting, and magnetic properties of substances at high pressures"; (5) Dizhur E M, Venttsel V A and Voronovskii A N (Institute for High Pressure Physics, RAS, Troitsk, Moscow region), "Quantum transport at high pressures"; (6) Ryzhov V N, Barabanov A F, Magnitskaya M V and Tareyeva E E (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "Theoretical studies of condensed matter"; (7) Bugakov V I, Antanovich A A, Konyaev Yu S and Slesarev V N (Institute for High Pressure Physics, RAS, Troitsk, Moscow region) "Designing new construction and superhard materials and related tools." An abridged version of reports 1 -6 is presented below. • The Institute for High Pressure Physics is now 50 (opening address), Stishov S M Physics-Uspekhi, 2008, Volume 51, Number 10, Pages 1055-1059 • Large-volume high-pressure devices for physical investigations, Khvostantsev L G and Slesarev V N Physics-Uspekhi, 2008, Volume 51, Number 10, Pages 1059

  12. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    SciTech Connect

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  13. Pressurized pyrolysis of dried distillers grains with solubles and canola seed press cake in a fixed-bed reactor.

    PubMed

    Ateş, Funda; Miskolczi, Norbert; Saricaoğlu, Beyza

    2015-02-01

    Pressurized pyrolysis of biomasses was carried in a fixed bed reactor to obtain gases, bio-oils and chars at elevated temperatures. The products were characterized by GC-MS, FTIR, viscometer, SEM, BET and EDXRFS methods. Experiments were performed at 1, 5 and 10 bar pressure and 400, 500 and 600°C temperatures. The experimental results show that in all the experimental condition the yield of bio-oil from DDGS as higher than that of canola. Yield of non-condensable gases and chars increased, while that of liquid products decreased by pressure. Increasing pressure favoured the formation of low molecular weight gas, such as H2. Maximum surface area of chars was obtained at atmospheric pressure and the surface areas decreased rapidly with increasing pressure. GC/MS results shows that the amount of fatty acids in bio-oils was increased by increasing pressure and bio-oils showed non-Newtonian behavior. Based on EDXRFS results, bio-oils and char contained lots of elements. PMID:25484126

  14. Buoyancy induced limits for nanoparticle synthesis experiments in horizontal premixed low-pressure flat-flame reactors

    NASA Astrophysics Data System (ADS)

    Weise, C.; Faccinetto, A.; Kluge, S.; Kasper, T.; Wiggers, H.; Schulz, C.; Wlokas, I.; Kempf, A.

    2013-06-01

    Premixed low-pressure flat-flame reactors can be used to investigate the synthesis of nanoparticles. The present work examines the flow field inside such a reactor during the formation of carbon (soot) and iron oxide (from Fe(CO)5) nanoparticles, and how it affects the measurements of nanoparticle size distribution. The symmetry of the flow and the impact of buoyancy were analysed by three-dimensional simulations and the nanoparticle size distribution was obtained by particle mass spectrometry (PMS) via molecular beam sampling at different distances from the burner. The PMS measurements showed a striking, sudden increase in particle size at a critical distance from the burner, which could be explained by the flow field predicted in the simulations. The simulation results illustrate different fluid mechanical phenomena which have caused this sudden rise in the measured particle growth. Up to the critical distance, buoyancy does not affect the flow, and an (almost) linear growth is observed in the PMS experiments. Downstream of this critical distance, buoyancy deflects the hot gas stream and leads to an asymmetric flow field with strong recirculation. These recirculation zones increase the particle residence time, inducing very large particle sizes as measured by PMS. This deviation from the assumed symmetric, one-dimensional flow field prevents the correct interpretation of the PMS results. To overcome this problem, modifications to the reactor were investigated; their suitability to reduce the flow asymmetry was analysed. Furthermore, 'safe' operating conditions were identified for which accurate measurements are feasible in premixed low-pressure flat-flame reactors that are transferrable to other experiments in this type of reactor. The present work supports experimentalists to find the best setup and operating conditions for their purpose.

  15. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system.

    PubMed

    Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

    2013-01-01

    Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ(13)Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate. PMID:24348470

  16. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

    PubMed Central

    Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

    2013-01-01

    Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0–100 MPa) and temperature (0–70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate. PMID

  17. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    NASA Astrophysics Data System (ADS)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  18. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  19. Russian Grouting Experience

    SciTech Connect

    Langton, C.A.

    2002-10-15

    The final report on Russian Grouting experience provided an opportunity for international cooperation and access to Russian grouting/waste form experience. the data on radiolytic gas generation from grout mixtures was already used in evaluation of the source of hydrogen and methane generation detected in the sampling ports around the SRS high-level waste tanks in 2002. The concept of venting the radiolytic gases from a waste form by adding porous aggregate is being considered for future cement-based TRU waste forms at SRS. The objectives of this work were to document the Russian experience on grouting for waste forms and tank closures or other decommissioning applications.

  20. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  1. Research on Russian National Character

    ERIC Educational Resources Information Center

    Na, Zhuo

    2008-01-01

    The special geographical location Russia lies in creates the unique character of the Russian nation. Based on the dual nature of the Russian national character, the Russian geographical environment and the analysis of its social structure, this text tries to explore the reasons of the dual nature of Russian national character.

  2. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    SciTech Connect

    Edwards, D.J.; Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  3. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    SciTech Connect

    Gentry, Cole A; Godfrey, Andrew T; Terrani, Kurt A; Gehin, Jess C; Powers, Jeffrey J; Maldonado, G Ivan

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  4. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    DOE PAGESBeta

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore » fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the

  5. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    SciTech Connect

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

    A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to

  6. Correlation of streamer current pulses associated with adjacent high voltage needles in atmospheric pressure cold plasma reactors

    NASA Astrophysics Data System (ADS)

    Wemlinger, Erik; Pedrow, Patrick

    2011-10-01

    We hypothesize that for a 12 needle array in an atmospheric pressure cold plasma reactor there will be correlation between needle corona current pulses. Guaitella et al. have shown in their surface dielectric barrier discharge that synchronous surface streamers are likely triggered by photodesorbed negative charges with binding energy (at the surface of the dielectric) less than 3.5 eV. The reactor used in our work has two rings of axially aligned needles. The current in each needle is measured with broad band current sensors that respond primarily to free electron drift. Digital signal processing will be used to analyze correlation between streamer current pulses. A 60 Hz 10 kVRMS voltage source produces the streamers and concomitantly the cold plasma. The current pulse correlation will be studied between 1 needle and each of the other 11 needles with the expectation that nearest neighbor needles will have the highest correlation. Understanding correlated streamer current pulses will inform reactor modeling and reactor optimization. O. Guaitella, I. Marinov, A. Rousseau, Applied Physics Letters, 98, 2011.

  7. Features of temperature control of fuel element cladding for pressurized water nuclear reactor ``WWER-1000'' while simulating reactor accidents

    NASA Astrophysics Data System (ADS)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-01

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones "cladding-junction" and "junction-coolant". The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ˜5% of maximum value by the final moment of the stage of linear increase in the temperature.

  8. Station Tour: Russian Segment

    NASA Video Gallery

    Expedition 33 Commander Suni Williams concludes her tour of the International Space Station with a visit to the Russian segment, which includes Zarya, the first segment of the station launched in 1...

  9. Save Russian science

    NASA Astrophysics Data System (ADS)

    Feigel'Man, Mikhail

    2007-03-01

    Despite 15 years of turbulent change, 'brain drain' and a shortage of research funds, Russian science has survived, although in a much diminished state. International investment and collaboration over the next ten years could bring it back from the brink.

  10. Russian Contract Procurement Document

    SciTech Connect

    Tobin, J G

    2010-03-29

    This contract supports the enhancement of physical protection or nuclear material control and accounting systems at institutes or enterprises of the newly independent states under the material protection control and accounting (MPC&A) program. The contract is entered into pursuant to the MPC&A Program, a gratuitous technical assistance program, in accordance with the bilateral Agreements between the Russian Federation and the United States of America concerning the Safe and Secure Transportation, Storage and Destruction of Weapons and the Prevention of Weapons Proliferation of June 1992, as extended and amended by Protocol signed of June 1999, Agreement between the Government of the Russian Federation regarding Cooperation in the Area of Nuclear Materials Physical Protection, Control and Accounting of October 1999 and the Russian Federation law of May 1999 on the taxation exemption of gratuitous technical assistance with Russian Federation under registration No.DOE001000.

  11. The Low Temperature Oxidation of 2,7-Dimethyloctane in a Pressurized Flow Reactor

    NASA Astrophysics Data System (ADS)

    Farid, Farinaz

    The complexity of real fuels has fostered the use of simple mixtures of hydrocarbons whose combustion behavior approximates that of real fuels in both experimental and computational studies to develop models of the combustion of the real fuel. These simple mixtures have been called surrogates. Lightly branched paraffins are an important class of constituents in gasoline, diesel and aviation turbine fuels and therefore are primary candidates for use as a component in a surrogate. Unfortunately, fundamental studies on combustion characteristics of high molecular weight mono- and di-methylated iso-paraffins are scarce. Therefore, this study was designed to investigate the low-temperature oxidation of 2,7-dimethyloctane (2,7-DMO) (C10H22), a lightly branched isomer of decane. Replicate 2,7-DMO oxidation experiments were conducted in a pressurized flow reactor (PFR) over the temperature range of 550 -- 850 K, at a pressure of 8 atm and an equivalence ratio of 0.3 in 4.21% oxygen / nitrogen. The reactivity was mapped by continuous monitoring of CO, CO 2, and O2 using a non-dispersive infrared (NDIR) carbon monoxide / carbon dioxide analyzer and an electrochemical oxygen sensor. For examining the underlying reaction chemistry, detailed speciation of samples was performed at selected temperatures using a gas chromatograph with a flame ionization detector coupled to a mass spectrometer. Comparable oxidation experiments for n-decane were carried out to examine the unique effects of branching on fuel reactivity and distribution of major stable intermediates. For both isomers, the onset of negative temperature coefficient (NTC) region was observed near 700 K, with the reactivity decreasing with increasing the temperature. The flow reactor study of n-decane oxidation confirmed that the isomerization reduces the amount of CO produced at peak reactivity. In addition to reaction inhibition, branching affected the distribution of C2-C 4 olefin intermediates. While the oxidation of

  12. Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

    SciTech Connect

    Yun, D.; Taiwo, T. A.; Kim, T. K.; Mohamed, A.; Nuclear Engineering Division

    2010-10-01

    The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

  13. Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

    SciTech Connect

    Hoye, Robert L. Z. E-mail: jld35@cam.ac.uk; MacManus-Driscoll, Judith L. E-mail: jld35@cam.ac.uk; Muñoz-Rojas, David; Nelson, Shelby F.; Illiberi, Andrea; Poodt, Paul

    2015-04-01

    Atmospheric pressure spatial atomic layer deposition (AP-SALD) has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

  14. Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

    SciTech Connect

    Pennell, W.E.; Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Merkle, J.G.

    1996-04-01

    Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K{sub Jc} data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K{sub Ic} data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K{sub Ic} and K{sub IR} curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested in the lower transition temperature range was shown to be strain controlled. A strain-based dual-parameter fracture-toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture.

  15. Modeling the shear rate and pressure drop in a hydrodynamic cavitation reactor with experimental validation based on KI decomposition studies.

    PubMed

    Badve, Mandar P; Alpar, Tibor; Pandit, Aniruddha B; Gogate, Parag R; Csoka, Levente

    2015-01-01

    A mathematical model describing the shear rate and pressure variation in a complex flow field created in a hydrodynamic cavitation reactor (stator and rotor assembly) has been depicted in the present study. The design of the reactor is such that the rotor is provided with surface indentations and cavitational events are expected to occur on the surface of the rotor as well as within the indentations. The flow characteristics of the fluid have been investigated on the basis of high accuracy compact difference schemes and Navier-Stokes method. The evolution of streamlining structures during rotation, pressure field and shear rate of a Newtonian fluid flow have been numerically established. The simulation results suggest that the characteristics of shear rate and pressure area are quite different based on the magnitude of the rotation velocity of the rotor. It was observed that area of the high shear zone at the indentation leading edge shrinks with an increase in the rotational speed of the rotor, although the magnitude of the shear rate increases linearly. It is therefore concluded that higher rotational speeds of the rotor, tends to stabilize the flow, which in turn results into less cavitational activity compared to that observed around 2200-2500RPM. Experiments were carried out with initial concentration of KI as 2000ppm. Maximum of 50ppm of iodine liberation was observed at 2200RPM. Experimental as well as simulation results indicate that the maximum cavitational activity can be seen when rotation speed is around 2200-2500RPM. PMID:24924259

  16. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    SciTech Connect

    Ren, Weiju; Terry, Totemeier

    2006-10-01

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

  17. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    SciTech Connect

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  18. Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

    NASA Astrophysics Data System (ADS)

    Han, Guangdong; Lu, Zhanpeng; Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu

    2015-12-01

    The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen.

  19. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    SciTech Connect

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  20. Gas-phase chemistry during the conversion of cyclohexane to carbon: Flow reactor studies at low and intermediate pressure

    SciTech Connect

    Osterheld, T.H.; Allendorf, M.D.; Larson, R.

    1995-07-01

    The gas-phase branching during the conversion of cyclohexane to solid carbon has been measured in a high-temperature-flow reactor. The experiments show that cyclohexane decomposes into a broad distribution of hydrocarbons that further decompose into the more kinetically stable products hydrogen, methane, acetylene, ethylene, benzene, and PAH. At 1363 K, the evolution to these species occurs quickly. We also observe the buildup of significant amounts of aromatic molecules at later stages in the decomposition, with as much as 15% of the total carbon in PAH and 25% in benzene. At later stages, the gas-phase molecules react slowly, even though the system is not at equilibrium, because of their kinetic stability and the smaller radical pool. The decomposition does not appear to depend sensitively on pressure in the regime of 25 to 250 torr. Thus, to a first approximation, these results can be extrapolated to atmospheric pressure.

  1. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  2. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

    NASA Astrophysics Data System (ADS)

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  3. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

    SciTech Connect

    Oßwald, Patrick; Köhler, Markus

    2015-10-15

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  4. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems.

    PubMed

    Oßwald, Patrick; Köhler, Markus

    2015-10-01

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes. PMID:26520986

  5. Slurry pumping techniques for feeding high-pressure coal gasification reactors

    NASA Technical Reports Server (NTRS)

    Bair, W. G.; Tarman, P. B.

    1977-01-01

    Operating experience in pumping coal and coal char slurries at pressures up to 1500 psig is discussed. The design specifications for the mixing tanks, pumps, piping, and slurry heaters are given along with pressure drop and minimum flow velocity data on water-lignite slurries.

  6. The Russian Virtual Observatory

    NASA Astrophysics Data System (ADS)

    Dluzhnevskaya, O. B.; Malkov, O. Yu.; Kilpio, A. A.; Kilpio, E. Yu.; Kovaleva, D. A.; Sat, L. A.

    The Russian Virtual Observatory (RVO) will be an integral component of the International Virtual Observatory (IVO). The RVO has the main goal of integrating resources of astronomical data accumulated in Russian observatories and institutions (databases, archives, digitized glass libraries, bibliographic data, a remote access system to information and technical resources of telescopes etc.), and providing transparent access for scientific and educational purposes to the distributed information and data services that comprise its content. Another goal of the RVO is to provide Russian astronomers with on-line access to the rich volumes of data and metadata that have been, and will continue to be, produced by astronomical survey projects. Centre for Astronomical Data (CAD), among other Russian institutions, has had the greatest experience in collecting and distributing astronomical data for more than 20 years. Some hundreds of catalogs and journal tables are currently available from the CAD repository. More recently, mirrors of main astronomical data resources (VizieR, ADS, etc) are now maintained in CAD. Besides, CAD accumulates and makes available for the astronomical community information on principal Russian astronomical resources.

  7. The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

    NASA Astrophysics Data System (ADS)

    Herbschleb, C. T.; van der Tuijn, P. C.; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; van Spronsen, M. A.; Bergman, M.; Crama, L.; Taminiau, I.; Ofitserov, A.; van Baarle, G. J. C.; Frenken, J. W. M.

    2014-08-01

    To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

  8. Changes in magnetic parameters of neutron irradiated SA 508 Cl. 3 reactor pressure vessel forging and weld surveillance specimens

    NASA Astrophysics Data System (ADS)

    Chi, Se-Hwan; Chang, Kee-Ok; Hong, Jun-Hwa; Kuk, Il-Hiun; Kim, Chong-Oh

    1999-04-01

    Irradiation-induced changes in the magnetic parameters and mechanical properties were measured and compared to explore possible correlations for reactor pressure vessel (RPV) forging and weld surveillance Charpy specimens which were irradiated to the neutron fluence of 2.3×1019n/cm2 (E>1.0 MeV) in a typical pressurized water reactor environment at 290 °C. For mechanical property parameters, Vickers microhardness, tensile and Charpy impact tests were performed and saturation magnetization (Ms), remanence (Mr), coercivity (Hc), and Barkhausen noise amplitude (BNA) were measured for magnetic parameters for both unirradiated and irradiated specimens, respectively. Results of mechanical property measurements showed an increase in yield and tensile strength, Vickers microhardness, 30 ft. lb indexed RTNDT and a decrease in Charpy upper-shelf energy irrespective of forging and weld metals. Hysteresis loops appeared to turn clockwise, resulting in an increase in Hc, and BNA appeared to decrease after irradiation. Both magnetic parameters showed viable correlations to the changes in mechanical parameters (Vickers microhardness, Charpy upper shelf energy) due to irradiation. Even limited, the present study seems to show additional possibilities for the application of this magnetic method in monitoring the mechanical parameter changes due to neutron irradiation.

  9. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    SciTech Connect

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  10. The ReactorSTM: Atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions

    SciTech Connect

    Herbschleb, C. T.; Tuijn, P. C. van der; Roobol, S. B.; Navarro, V.; Bakker, J. W.; Liu, Q.; Stoltz, D.; Cañas-Ventura, M. E.; Verdoes, G.; Spronsen, M. A. van; Bergman, M.; Crama, L.; Taminiau, I.; Frenken, J. W. M.; Ofitserov, A.; Baarle, G. J. C. van

    2014-08-15

    To enable atomic-scale observations of model catalysts under conditions approaching those used by the chemical industry, we have developed a second generation, high-pressure, high-temperature scanning tunneling microscope (STM): the ReactorSTM. It consists of a compact STM scanner, of which the tip extends into a 0.5 ml reactor flow-cell, that is housed in a ultra-high vacuum (UHV) system. The STM can be operated from UHV to 6 bars and from room temperature up to 600 K. A gas mixing and analysis system optimized for fast response times allows us to directly correlate the surface structure observed by STM with reactivity measurements from a mass spectrometer. The in situ STM experiments can be combined with ex situ UHV sample preparation and analysis techniques, including ion bombardment, thin film deposition, low-energy electron diffraction and x-ray photoelectron spectroscopy. The performance of the instrument is demonstrated by atomically resolved images of Au(111) and atom-row resolution on Pt(110), both under high-pressure and high-temperature conditions.

  11. Neutron dosimetry in the containment of a pressurized water reactor using a neutron-sensitive beta/gamma dosimetry system

    SciTech Connect

    Kralick, S.C.; Watson, J.E. Jr.; Croslin, S.W.

    1986-06-01

    In this study the Panasonic UD-802 dosimeter was evaluated as a potential neutron dosimeter for use in the containment of a pressurized water reactor by comparing the results from the UD-802 with remmeter readings. The Panasonic UD-802 dosimeter is used routinely as a beta and gamma dosimeter but due to the natural Li and B in the thermoluminescent materials, it is also sensitive to neutrons. Since a dosimeter's response to neutrons is energy-dependent, proper calibration of the UD-802 in the environment for which it is to be used was an important consideration of the study. To calibrate the system, UD-802 dosimeters were mounted on polyethylene phantoms and irradiated to reference doses at selected locations in containment. The reference doses were determined based on remmeter dose-rate measurements and stay times. The thermoluminescent response of the dosimeters and the reference measurements were used to obtain a response ratio at each location. The average response ratio (unit of dosimeter response per millirem) was 3.7 and all response ratios were within +/-30% of this mean value. Specific characteristics of the UD-802 were also investigated, that is, the effects that dosimeter distance from the phantom and a person's movement through containment have on response. The dosimeter distance from the phantom was found to have a minimal effect on response, but the system was found to be dependent upon the angle of the phantom relative to the reactor core, necessitating a correction in the calibration factor. The overall conclusion of this study was that the Panasonic UD-802 dosimeter can be used for neutron dosimetry in containment of a pressurized water reactor.

  12. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  13. Pressure-drops control strategy in a fixed-bed reactor.

    PubMed

    Thalasso, F; Razo-Flores, E; Ancia, R; Naveau, H P; Nyns, E J

    2001-01-29

    This paper presents a strategy to control pressure-drops (head loss) in a biofilter designed according to the "Mist-Foam" concept. This concept is based on the mixing of the gaseous substrate and a liquid nutrient solution with an atomization nozzle to generate a mist passing subsequently through a synthetic polyurethane foam. In this type of bioreactor, the microbial growth reduces progressively the empty bed volume of the biofilter and causes an increase in the pressure-drops. This phenomenon can result in a complete clogging of the biofilter. The strategy of pressure-drops control presented here consists of successive interruption of the liquid flow, automatically controlled, resulting in a drying effect of the biomass. Tested during a 160 days experiment, this system has permitted to reduce and stabilize the pressure-drops in a biofilter in which the carrier exhibited a high likelihood of clogging. PMID:11118687

  14. Evaluation of Nodal Reactor Physics Methods for Quasi-Static and Time-Dependent Coupled Neutronic Thermal - Analysis of Pressurized Water Reactor Transients

    NASA Astrophysics Data System (ADS)

    Feltus, Madeline Anne

    1990-01-01

    This thesis examines coupled time-dependent thermal -hydraulic (T/H) and neutronics solution methods for Pressurized Water Reactor (PWR) transient analysis. The degree of equivalence is evaluated between the typical quasi-static approach and a newly-developed iterative tandem method. Four specific PWR transients that exhibit a wide range of Reactor Coolant System (RCS) T/H response were investigated: (1) a Station Blackout Anticipated Transient Without Scram (ATWS), (2) a Loss of Feedwater ATWS, (3) a Total Loss of RCS Flow with Scram, and (4) a Main Steam Line Break (MSLB). Rather than using simplified RCS and core models, the theory and method in this thesis were applied practically by using realistic models for an actual four-loop Westinghouse PWR plant. The time-dependent STAR kinetics code, based on the QUANDRY Analytic Nodal Method, and the RETRAN and MCPWR T/H systems codes were used to develop a new, fully coupled, tandem STAR/MCPWRQ methodology that runs tandemly on an enhanced 386/387 IBM PC architecture. MCPWRQ uses externally calculated power input rather than point kinetics power level results. The tandem method was compared to quasi -static STAR and time-dependent STAR 2-D and 3-D kinetics results. The new STAR/MCPWRQ method uses RETRAN time-dependent T/H and point kinetics power input as a first estimate. STAR and MCPWRQ are used tandemly to couple STAR 3-D, time-dependent core power results with the MCPWRQ RCS T/H phenomena. This thesis shows that: (a) quasi-static and point kinetics methods are not able to describe severe PWR transient phenomena adequately; and (b) fully coupled, 3-D, time -dependent, tandem (or possibly parallel) analysis methods should be used for PWR reactor transients instead. By tandemly coupling the RCS response in terms of updated core inlet conditions with 3-D time-dependent core kinetics response, the core power response and T/H conditions are forced to be self-consistent during the entire transient. The transient analyses

  15. Studying Russian and Soviet History.

    ERIC Educational Resources Information Center

    Ascher, Abraham, Ed.

    These essays were written to assist teachers in the task of making Russian history intelligible to young U.S. students. In "An Approach to Russian History," Edward Keenan proposes that students need to gain a better understanding of how Russians perceive themselves and their history. In "Pre-Petrine Russia," Andrzej S. Kaminski focuses on the…

  16. Optimizing the water chemistry of the Novovoronezh NPP Units 3 and 4 to prevent growth of coolant pressure difference across the reactor

    NASA Astrophysics Data System (ADS)

    Kritskii, V. G.; Rodionov, Yu. A.; Berezina, I. G.; Zelenina, E. V.; Gavrilov, A. V.; Vitkovskii, S. L.; Shchedrin, M. G.; Galanin, A. V.; Gorburov, V. I.

    2014-02-01

    For a few fuel campaigns in the first years of their extended service life, the Novovoronezh NPP Units 3 and 4 had to operate with a limited reactor thermal power output due to constraints imposed on certain parameters of the reactor core. The decrease of thermal power output during a campaign was accompanied by a growth of pressure difference across the reactor caused by accumulation of deposits in the fuel assemblies, which in turn led to a growth in the number of leaky fuel rods. Compensating measures were implemented based on the results obtained from the performed investigations and modeling of corrosion product mass transfer processes, which in the final analysis made it possible to minimize the accumulation of deposits in fuel assemblies and loss of fuel rod tightness, and to exclude the need to limit the reactor thermal power as demanded by constraints imposed on the reactor core parameters.

  17. A simple kinetic analysis of syngas during steam hydrogasification of biomass using a novel inverted batch reactor with instant high pressure feeding.

    PubMed

    Fan, Xin; Liu, Zhongzhe; Norbeck, Joseph M; Park, Chan S

    2016-01-01

    A newly designed inverted batch reactor equipped with a pressure-driven feeding system was built for investigating the kinetics of syngas during the steam hydrogasification (SHR) of biomass. The system could instantly load the feedstock into the reactor at high temperature and pressure, which simulated the way to transport the feedstock into a hot and pressurized gasifier. Experiments were conducted from 600°C to 700°C. The inverted reactor showed very high heating rate by enhancing the carbon conversion and syngas production. The kinetic study showed that the rates of CH4, CO and CO2 formation during SHR were increased when the gasification temperature went up. SHR had comparatively lower activation energy for CH4 production. The activation energies of CH4, CO and CO2 during SHR were 42.8, 51.8 and 14kJ/mol, respectively. PMID:26562689

  18. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 5 - North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.

    1993-01-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor (AFR) criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial pressurized-water reactors (PWR). The analysis methodology selected for all calculations reported herein was the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies in the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and to provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. This unit and cycle were chosen for a previous analysis using a different methodology because detailed isotopics from multidimensional reactor calculations were available from the Virginia Power Company. These data permitted comparison of criticality calculations directly using the utility-calculated isotopics to those using the isotopics generated by the SCALE-4 SAS2H

  19. Teaching Russian Civilization.

    ERIC Educational Resources Information Center

    Vakar, Nicholas

    1949-01-01

    Prior to World War Two, Slavic studies in America treated history, literature, and language as isolated disciplines and often neglected the study of Russian literature written after 1917. The pragmatic necessities of the war questioned the relevance of this traditional approach and specialists appeared, concentrating their efforts on the recent…

  20. Russian Librarianship after Perestroika.

    ERIC Educational Resources Information Center

    Raymond, Boris

    1995-01-01

    Provides historical background and describes the change in the role of Russian libraries from conduits for party propaganda to organizations designed to serve the needs of their users. Considers problems for library educators in providing librarians with the skills needed in the new political and technological environment of post-Communist Russia.…

  1. Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1

    SciTech Connect

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1985-04-01

    This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

  2. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    SciTech Connect

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A.

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  3. Adolescent Russian roulette deaths.

    PubMed

    Collins, Kim A

    2010-03-01

    Adolescence, between the ages of 10 and 19 years, is a unique period both physically and emotionally. During this time of life, individuals are known to experiment and engage in risky behavior, sometimes with unforeseen morbidity and mortality. We also see suicide emerge as a manner of death in this age group. The most common method is gunshot wound and sometimes in the form of Russian roulette. Few studies have looked at deaths by Russian roulette, the victims, and scenarios. In particular, no study examines the adolescent victim of Russian roulette. To better understand and classify this entity, adolescent Russian roulette autopsy cases over a 20-year period were examined looking at the victims, scenarios, autopsy findings, cause and manner of death, and the weapons. All victims were males, ages 13 to 19 years, with a Black-to-White ratio of 1:1. No victim had a previous psychiatric history. Toxicology was positive for alcohol and/or marijuana in 50% of the victims. Friends were present when the victim shot himself which occurred in the home the majority of the time. In all but 1 case, premeditation of the game was involved as the victim provided the weapon for the roulette. The cause of death was gunshot wound to the head (6 to the right side, 1 to the mouth, 1 to the forehead), and the manner of death was suicide in 6 cases and accident in 2 cases. A review of the literature discusses the adolescent victim, suicide, and Russian roulette. PMID:20010290

  4. Russian Basic Course. Naval Terminology Dictionary: Russian-English, English-Russian.

    ERIC Educational Resources Information Center

    Defense Language Inst., Washington, DC.

    This dictionary is a companion to "Naval Terminology," developed by the Defense Language Institute and printed in 1971. Cyrillic script is used for the Russian terminology. Two main sections define naval terms from Russian to English and from English to Russian. A list of abbreviations concludes the text. (RL)

  5. Numerical investigation of the film uniformity during the surface coating of charged nanoparticles in a low pressure plasma reactor

    NASA Astrophysics Data System (ADS)

    Pourali, N.; Foroutan, G.

    2016-07-01

    The uniformity of film deposition on charged nanoparticles, trapped near the sheath of a capacitively coupled plasma reactor, is studied by numerical simulation of the multi-fluid plasma equations, surface deposition processes, and nanoparticle heating effects. It is found that the anisotropy in the ion flux onto the powered electrode may be hold responsible for the film nonuniformity. The nonuniformity increases with increasing of the particle radius, although small particles lose sphericity faster than the large particles. Because of the electron temperature dependence of the deposition rate and the incident ion flux, higher electron temperatures lead to more nonuniform film deposition. However, the uniformity is improved and the sphericity is restored by the increase in the background gas pressure and/or temperature.

  6. Analysis of proposed gamma-ray detection system for the monitoring of core water inventory in a pressurized water reactor

    SciTech Connect

    Markoff, D.M.

    1987-12-01

    An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and downcomer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport path, and the relative void fraction in different water regions) on the detector response are studied. The calculational results have been validated by a favorable comparison with LOFT experimental data. Within the limitations and approximations considered in the analysis, the results indicate that the gamma-ray detection scheme is able to unambiguously respond to changes in the coolant inventory within any vessel water region.

  7. Effects of control system failures on transients and accidents at a 3-loop Westinghouse pressurized water reactor. Volume 2. Appendices

    SciTech Connect

    Bruske, S.J.; Davis, C.B.; Ogden, D.M.; Ransom, C.B.; Stitt, B.D.; Stromberg, H.M.; Waterman, M.E.

    1985-10-01

    Safety Implications of Control Systems (A-47) was approved as an Unresolved Safety Issue (USI) by the Nuclear Regulatory Commission (NRC) in December of 1980. USI A-47 is concerned with the potential for transients or accidents being made more severe than previously analyzed as a result of control system failures. This report describes the work performed on the effects of control system failures on transients and accidents at a Westinghouse 3-loop pressurized water reactor. In this volume, the appendices contain detailed information consisting of the FMEA (failure mode and analysis) results, an in-depth description of the computer model, the deterministic computer analyses, and responses to comments made by Carolina Power and Light Company and Westinghouse Electric Corporation.

  8. Stability of cement-glass packages containing sodium borate salt generated from pressurized water reactor power plants

    SciTech Connect

    Izumida, T.; Kawamura, F.; Chino, K.; Kikuchi, M.

    1987-08-01

    A new solidification technique using cement-glass, which is a mixture of sodium silicate and cement, was studied for solidification of sodium borate salt of liquid waste generated from pressurized water reactor plants. When the sodium borate salt was solidified with the cement-glass, the resulting package contained eight times more sodium borate than was found in cement because it did not interact with sodium borate. The leaching ratio of cesium ion from the cement-glass package was one-tenth that of cement. Its low leaching ratio was due to the high cesium adsorption ability of cement-glass. The ratio could be theoretically evaluated by considering the cesium adsorption-desorption equilibrium.

  9. COMSOL Modeling of Transport of Neutral Radicals to Substrate Surfaces Located Downstream from an Atmospheric Pressure Weakly Ionized Plasma Reactor

    NASA Astrophysics Data System (ADS)

    Islam, Rokibul; Lekobou, William; Wemlinger, Erik; Pedrow, Patrick

    2012-10-01

    An Atmospheric Pressure Weakly Ionized Plasma (APWIP) Reactor generates a significant number of charged particles and neutral radicals. In our work the carrier gas is argon and the precursor molecule is acetylene. The APWIP is generated by corona discharges associated with an array of high voltage metal needles facing a grounded metal screen. Neutral radical transport downstream from the grounded screen to the substrate via diffusion and convection will be modeled with COMSOL, a finite element software package. Substrates will include objects with various shapes and characteristic dimensions that range from nanometers to centimeters. After the model is validated against canonical problems with known solutions, thin film deposition rates will be compared with experimentally measured results. Substrate geometries will include discs, spheres, fibers and highly porous surfaces such as those found on asphalt road surfaces. A single generic neutral radical will be used to represent the entire family of neutral radicals resulting from acetylene bond scission by free electron impact.

  10. Magnetic evaluation of irradiation hardening in A533B reactor pressure vessel steels: Magnetic hysteresis measurements and the model analysis

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Yamamoto, T.; Klingensmith, D.; Odette, G. R.; Kikuchi, H.; Kamada, Y.

    2012-03-01

    We report results of measurements of magnetic minor hysteresis loops for neutron-irradiated A533B nuclear reactor pressure vessel steels varying alloy composition and irradiation condition. A minor-loop coefficient, which is obtained from a scaling power law between minor-loop parameters exhibits a steep decrease just after irradiation, followed by a maximum in the intermediate fluence regime for most alloys. A model analysis assuming Avrami-type growth for Cu-rich precipitates and an empirical logarithmic law for relaxation of residual stress demonstrates that an increment of the coefficient due to Cu-rich precipitates increases with Cu and Ni contents and is in proportion to a yield stress change, which is related to irradiation hardening.

  11. On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Meslin, E.; Malerba, L.; Hernández-Mayoral, M.; Bergner, F.; Pareige, P.; Radiguet, B.; Almazouzi, A.

    2010-11-01

    A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ˜7 × 10 19 n cm -2. This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.

  12. Thermal annealing of the reactor pressure vessel NPP Unit 2 in Jaslovske Bohunice for its radiation embrittlement regeneration

    SciTech Connect

    Kupca, L.; Cepcek, S.

    1993-12-01

    The status of the preparation works for the thermal annealing operation at reactor pressure vessel (RPV) V-230-type Unit 2 in Jaslovske Bohunice planned for Autumn 1993 is presented in this paper. The producer of the RPV W-213 type, SKODA Works, will perform the thermal annealing operation and manufacture all equipment needed. During the planned shutdown for the refueling operation of this unit in September 1989, samples were prepared from base material (BM) and weld metal (WM) by means of special equipment used for the analysis of the chemical composition in the Nuclear Power Plants Research Institute (VUJE) laboratories. Results of the analysis of the irradiated samples and the hardness measurements of RPV material before and after annealing operation serves as the measure of radiation embrittlement recovery efficiency. Possible extension of the operation life of RPVs of WWER type by means of suitable provisions during normal operation before thermal annealing is also discussed.

  13. Neutron spectrum effect on pressure vessel embrittlement: Dosimetry and qualification of irradiation locations in OSIRIS and SILOE reactors

    SciTech Connect

    Alberman, A.; Bourdet, L.; Carcreff, H.; Beretz, D.

    1994-12-31

    Two irradiation experiments have been undertaken in OSIRIS (Saclay) and SILOE (Grenoble) reactors, in order to establish the correlation between the embrittlement of pressure vessel steels and neutron spectrum. Target fluence is 0.1 dpa for both experiments. This damage fluence corresponds to a fluence of 7.5 10{sup 19} n.cm{sup {minus}2} E > 1 MeV (7.5 10{sup 15} n.m{sup {minus}2}) in the case of a well moderated light water spectrum, but only 45 10{sup 19} n.cm{sup {minus}2} in the case of the specially designed SILOE irradiation location. One irradiation run is now completed, the second one is underway. This paper presents the experimental dosimetry data and irradiation parameters obtained in the preliminary qualification program, needed to assess this damage correlation.

  14. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    SciTech Connect

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs.

  15. A kinetic model for impact/sliding wear of pressurized water reactor internal components: Application to rod cluster control assemblies

    SciTech Connect

    Zbinden, M.; Durbec, V.

    1996-12-01

    Certain internal components of Pressurized Water Reactors are damaged by wear when subjected to vibration induced by flow. In order to enable predictive calculation of such wear, one must have a model which takes account reliably of real damages. The modelling of wear represents a final link in a succession of numerical calculations which begins by the determination of hydraulic excitations induced by the flow. One proceeds, then, in the dynamic response calculation of the structure to finish up with an estimation of volumetric wear and of the depth of wear scars. A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which correspond to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work.

  16. Atmospheric pressure plasma chemical vapor deposition reactor for 100 mm wafers, optimized for minimum contamination at low gas flow rates

    NASA Astrophysics Data System (ADS)

    Anand, Venu; Nair, Aswathi R.; Shivashankar, S. A.; Mohan Rao, G.

    2015-08-01

    Gas discharge plasmas used for thinfilm deposition by plasma-enhanced chemical vapor deposition (PECVD) must be devoid of contaminants, like dust or active species which disturb the intended chemical reaction. In atmospheric pressure plasma systems employing an inert gas, the main source of such contamination is the residual air inside the system. To enable the construction of an atmospheric pressure plasma (APP) system with minimal contamination, we have carried out fluid dynamic simulation of the APP chamber into which an inert gas is injected at different mass flow rates. On the basis of the simulation results, we have designed and built a simple, scaled APP system, which is capable of holding a 100 mm substrate wafer, so that the presence of air (contamination) in the APP chamber is minimized with as low a flow rate of argon as possible. This is examined systematically by examining optical emission from the plasma as a function of inert gas flow rate. It is found that optical emission from the plasma shows the presence of atmospheric air, if the inlet argon flow rate is lowered below 300 sccm. That there is minimal contamination of the APP reactor built here, was verified by conducting an atmospheric pressure PECVD process under acetylene flow, combined with argon flow at 100 sccm and 500 sccm. The deposition of a polymer coating is confirmed by infrared spectroscopy. X-ray photoelectron spectroscopy shows that the polymer coating contains only 5% of oxygen, which is comparable to the oxygen content in polymer deposits obtained in low-pressure PECVD systems.

  17. Optimization of pressurized water reactor shuffling by simulated annealing with heuristics

    SciTech Connect

    Stevens, J.G.; Smith, K.S.; Rempe, K.R.; Downar, T.J.

    1995-09-01

    Simulated-annealing optimization of reactor core loading patterns is implemented with support for design heuristics during candidate pattern generation. The SIMAN optimization module uses the advanced nodal method of SIMULATE-3 and the full cross-section detail of CASMO-3 to evaluate accurately the neutronic performance of each candidate, resulting in high-quality patterns. The use of heuristics within simulated annealing is explored. Heuristics improve the consistency of optimization results for both fast- and slow-annealing runs with no penalty from the exclusion of unusual candidates. Thus, the heuristic application of designer judgment during automated pattern generation is shown to be effective. The capability of the SIMAN module to find and evaluate families of loading patterns that satisfy design constraints and have good objective performance within practical run times is demonstrated. The use of automated evaluations of successive cycles to explore multicycle effects of design decisions is discussed.

  18. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Sun, Mingyue; Hao, Luhan; Li, Shijian; Li, Dianzhong; Li, Yiyi

    2011-11-01

    Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  19. An experimental investigation of high pressure synthesis of diamond films using microwave plasma cavity reactor

    SciTech Connect

    Kuo, K.P.; Asmussen, J.

    1996-12-31

    Uniform diamond film deposition over 2 in. dia. Si substrates is experimentally investigated using a microwave plasma disk reactor and CH{sub 4}/H{sub 2} gas mixtures. The microwave plasma reactor is of the type developed at Michigan State University where the plasma is formed inside a 5 in. dia. disk-like discharge region located at one end of a cylindrical cavity applicator. The applicator employs the internal tuning, i.e., sliding short and probe variation, to match the input impedance of the plasma loaded applicator to the feed waveguide impedance. The deposited films were characterized by (1) measuring their uniformity over 2 in. dia., (2) growth rate in {micro}m/h and mg/h, (3) Raman Spectra, (4) film morphology and (5) film texture. Film growth rate, morphology, texture and Raman Spectra are recorded vs. CH{sub 4}/H{sub 2}, flow rate, substrate temperature and deposition time. Five hour experiments indicate that the average film growth rate varied from a few {micro}m/h with c = 1% to over 4.3 {micro}m/h at c = 3% and then decreased for high c ratios. Average growth rate also varies with flow rate with a maximum of over 5 {micro}m/h. The growth rate increases as deposition time increases and reaches a maximum growth rate of 6.3 {micro}m/h({approximately}44 mg/h) at t = 100 h. The film growth rate reaches a maximum around T{sub s} = 1,100 C for both 2% and 3% chemistries. 2 in. dia. free standing films were grown with thicknesses of 200--600 {micro}m and with uniformities of 10% over 2 in. dia. area.

  20. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    SciTech Connect

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  1. The ReactorAFM: Non-contact atomic force microscope operating under high-pressure and high-temperature catalytic conditions

    SciTech Connect

    Roobol, S. B.; Cañas-Ventura, M. E.; Bergman, M.; Spronsen, M. A. van; Onderwaater, W. G.; Tuijn, P. C. van der; Koehler, R.; Frenken, J. W. M.; Ofitserov, A.; Baarle, G. J. C. van

    2015-03-15

    An Atomic Force Microscope (AFM) has been integrated in a miniature high-pressure flow reactor for in-situ observations of heterogeneous catalytic reactions under conditions similar to those of industrial processes. The AFM can image model catalysts such as those consisting of metal nanoparticles on flat oxide supports in a gas atmosphere up to 6 bar and at a temperature up to 600 K, while the catalytic activity can be measured using mass spectrometry. The high-pressure reactor is placed inside an Ultrahigh Vacuum (UHV) system to supplement it with standard UHV sample preparation and characterization techniques. To demonstrate that this instrument successfully bridges both the pressure gap and the materials gap, images have been recorded of supported palladium nanoparticles catalyzing the oxidation of carbon monoxide under high-pressure, high-temperature conditions.

  2. The ReactorAFM: Non-contact atomic force microscope operating under high-pressure and high-temperature catalytic conditions

    NASA Astrophysics Data System (ADS)

    Roobol, S. B.; Cañas-Ventura, M. E.; Bergman, M.; van Spronsen, M. A.; Onderwaater, W. G.; van der Tuijn, P. C.; Koehler, R.; Ofitserov, A.; van Baarle, G. J. C.; Frenken, J. W. M.

    2015-03-01

    An Atomic Force Microscope (AFM) has been integrated in a miniature high-pressure flow reactor for in-situ observations of heterogeneous catalytic reactions under conditions similar to those of industrial processes. The AFM can image model catalysts such as those consisting of metal nanoparticles on flat oxide supports in a gas atmosphere up to 6 bar and at a temperature up to 600 K, while the catalytic activity can be measured using mass spectrometry. The high-pressure reactor is placed inside an Ultrahigh Vacuum (UHV) system to supplement it with standard UHV sample preparation and characterization techniques. To demonstrate that this instrument successfully bridges both the pressure gap and the materials gap, images have been recorded of supported palladium nanoparticles catalyzing the oxidation of carbon monoxide under high-pressure, high-temperature conditions.

  3. A 2-D Self-Consistent DSMC Model for Chemically Reacting Low Pressure Plasma Reactors

    SciTech Connect

    Bartel, Timothy J.; Economou, Demetre; Johannes, Justine E.

    1999-06-17

    This paper will focus on the methodology of using a 2D plasma Direct Simulation Monte Carlo technique to simulate the species transport in an inductively coupled, low pressure, chemically reacting plasma system. The pressure in these systems is typically less than 20 mtorr with plasma densities of approximately 10{sup 17} {number_sign}/m{sup 3} and an ionization level of only 0.1%. This low ionization level tightly couples the neutral, ion, and electron chemistries and interactions in a system where the flow is subsonic. We present our strategy and compare simulation results to experimental data for Cl{sub 2} in a Gaseous Electronics Conference (GEC) reference cell modified with an inductive coil.

  4. Rapid pressure swing absorption cleanup of post-shift reactor synthesis gas

    SciTech Connect

    Sirkar, K.K.; Majumdar, S.; Bhaumik, S. . Dept. of Chemical Engineering, Chemistry and Environmental Science)

    1993-01-29

    The theoretical model for the absorption part of a particular type of RAPSAB cycle (Mode 3) (see Technical Progress Report No. 7) has been developed. The numerical simulations of the model compare well with the experimental results presented in the last report (Technical Progress Report No. 7). A number of experiments were carried out also for Mode 2 type of operation by varying the time for initial pressurization of the hollow fiber module as well as the total absorption time. These were done to provide a basis for comparison with the theoretical model to be developed later. We have initiated RAPSAB studies with reactive absorbents such as 19.5 % aqueous solution of diethanolamine (DEA) for the absorption of C0[sub 2] from a C0[sub 2]-N[sub 2] mixture. Six experiments were carried out using Mode 3 type of operation and a C0[sub 2]-N[sub 2] mixture containing 9.9% CO, and balance N[sub 2]. Excellent purification was obtained. No C0[sub 2] was observed in the purified high pressure gas outlet for absorption time of up to 14 seconds; the purified high pressure gas flow rate was also considerable. Module No. 5 was used for all experiments. The details of the module are given in Technical Progress Report No. 7.

  5. Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 4-Three Mile Island Unit 1 Cycle 5

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The requirements of ANSI/ANS-8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original ''fresh'' composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using relevant and well-documented critical configurations from commercial pressurized water reactors. The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Isotopic densities for spent fuel assemblies in the core were calculated using the SCALE-4 SAS2H analytical sequence. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code family was used to extract the necessary isotopic densities from SAS2H results and to provide the data in the format required for SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) for the critical configuration. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all calculations. This volume of the report documents a reactor critical calculation for GPU Nuclear Corporation's Three Mile Island Unit 1 (TMI-1) during hot, zero-power startup testing for the beginning of cycle 5. This unit and cycle were selected because of their relevance in spent fuel benchmark applications: (1) cycle 5 startup occurred after an especially long downtime of 6.6 years; and (2) the core consisted primarily (75%) of burned fuel, with

  6. Numerical simulation of the nonlinear ultrasonic pressure wave propagation in a cavitating bubbly liquid inside a sonochemical reactor.

    PubMed

    Dogan, Hakan; Popov, Viktor

    2016-05-01

    We investigate the acoustic wave propagation in bubbly liquid inside a pilot sonochemical reactor which aims to produce antibacterial medical textile fabrics by coating the textile with ZnO or CuO nanoparticles. Computational models on acoustic propagation are developed in order to aid the design procedures. The acoustic pressure wave propagation in the sonoreactor is simulated by solving the Helmholtz equation using a meshless numerical method. The paper implements both the state-of-the-art linear model and a nonlinear wave propagation model recently introduced by Louisnard (2012), and presents a novel iterative solution procedure for the nonlinear propagation model which can be implemented using any numerical method and/or programming tool. Comparative results regarding both the linear and the nonlinear wave propagation are shown. Effects of bubble size distribution and bubble volume fraction on the acoustic wave propagation are discussed in detail. The simulations demonstrate that the nonlinear model successfully captures the realistic spatial distribution of the cavitation zones and the associated acoustic pressure amplitudes. PMID:26611813

  7. Russian Rocket Engine Test

    NASA Technical Reports Server (NTRS)

    1998-01-01

    NASA engineers successfully tested a Russian-built rocket engine on November 4, 1998 at the Marshall Space Flight Center (MSFC) Advanced Engine Test Facility, which had been used for testing the Saturn V F-1 engines and Space Shuttle Main engines. The MSFC was under a Space Act Agreement with Lockheed Martin Astronautics of Denver to provide a series of test firings of the Atlas III propulsion system configured with the Russian-designed RD-180 engine. The tests were designed to measure the performance of the Atlas III propulsion system, which included avionics and propellant tanks and lines, and how these components interacted with the RD-180 engine. The RD-180 is powered by kerosene and liquid oxygen, the same fuel mix used in Saturn rockets. The RD-180, the most powerful rocket engine tested at the MSFC since Saturn rocket tests in the 1960s, generated 860,000 pounds of thrust.

  8. Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

    SciTech Connect

    Dickson, T.L.

    1993-07-01

    Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140.

  9. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment

    NASA Astrophysics Data System (ADS)

    Zhang, Litao; Wang, Jianqiu

    2014-03-01

    Stress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics.

  10. Atmospheric pressure flow reactor: Gas phase chemical kinetics under tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L. (Inventor); Davis, Dennis D. (Inventor)

    1991-01-01

    A flow reactor for simulating the interaction in the troposphere is set forth. A first reactant mixed with a carrier gas is delivered from a pump and flows through a duct having louvers therein. The louvers straighten out the flow, reduce turbulence and provide laminar flow discharge from the duct. A second reactant delivered from a source through a pump is input into the flowing stream, the second reactant being diffused through a plurality of small diffusion tubes to avoid disturbing the laminar flow. The commingled first and second reactants in the carrier gas are then directed along an elongated duct where the walls are spaced away from the flow of reactants to avoid wall interference, disturbance or turbulence arising from the walls. A probe connected with a measuring device can be inserted through various sampling ports in the second duct to complete measurements of the first and second reactants and the product of their reaction at selected XYZ locations relative to the flowing system.

  11. Russian Rocket Engine Test

    NASA Technical Reports Server (NTRS)

    1998-01-01

    NASA engineers successfully tested a Russian-built rocket engine on November 4, 1998 at the Marshall Space Flight Center (MSFC) Advanced Engine Test Facility, which had been used for testing the Saturn V F-1 engines and Space Shuttle Main engines. The MSFC was under a Space Act Agreement with Lockheed Martin Astronautics of Denver to provide a series of test firings of the Atlas III propulsion system configured with the Russian-designed RD-180 engine. The tests were designed to measure the performance of the Atlas III propulsion system, which included avionics and propellant tanks and lines, and how these components interacted with the RD-180 engine. The RD-180 is powered by kerosene and liquid oxygen, the same fuel mix used in Saturn rockets. The RD-180, the most powerful rocket engine tested at the MSFC since Saturn rocket tests in the 1960s, generated 860,000 pounds of thrust. The test was the first test ever anywhere outside Russia of a Russian designed and built engine.

  12. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  13. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    SciTech Connect

    Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto; Wenner, Michael

    2012-07-01

    As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and

  14. Russian electrometallurgy: Achievements, problems, prospects

    NASA Astrophysics Data System (ADS)

    Utochkin, Yu. I.; Semin, A. E.

    2011-12-01

    The changes in the Russian metallurgy, in particular, electric furnace steelmaking, having occurred in the recent years are analyzed. The main increase in the steelmaking output is due to putting into operation of new electric furnaces in new miniworks and enterprises equipped earlier with open-hearth furnaces. Reaching the rated capacity of a furnace in Russia substantially lags behind foreign enterprises. Only 30-35% of the Russian market of corrosion-resistant steel are provided by Russian metal.

  15. Russian expats seek research reforms

    NASA Astrophysics Data System (ADS)

    Banks, Michael

    2009-11-01

    Over 170 Russian researchers working abroad have signed a letter addressed to the Russian president, Dmitry Medvedev, and prime minister Vladimir Putin raising concerns about "the catastrophic state of basic science" in the country. The letter, which appeared last month in the Moscow business paper Vedomosti, warns Russian leaders that unless urgent measures are implemented by the government, then science in the country may collapse.

  16. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  17. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    SciTech Connect

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  18. ELIXYS - a fully automated, three-reactor high-pressure radiosynthesizer for development and routine production of diverse PET tracers

    PubMed Central

    2013-01-01

    Background Automated radiosynthesizers are vital for routine production of positron-emission tomography tracers to minimize radiation exposure to operators and to ensure reproducible synthesis yields. The recent trend in the synthesizer industry towards the use of disposable kits aims to simplify setup and operation for the user, but often introduces several limitations related to temperature and chemical compatibility, thus requiring reoptimization of protocols developed on non-cassette-based systems. Radiochemists would benefit from a single hybrid system that provides tremendous flexibility for development and optimization of reaction conditions while also providing a pathway to simple, cassette-based production of diverse tracers. Methods We have designed, built, and tested an automated three-reactor radiosynthesizer (ELIXYS) to provide a flexible radiosynthesis platform suitable for both tracer development and routine production. The synthesizer is capable of performing high-pressure and high-temperature reactions by eliminating permanent tubing and valve connections to the reaction vessel. Each of the three movable reactors can seal against different locations on disposable cassettes to carry out different functions such as sealed reactions, evaporations, and reagent addition. A reagent and gas handling robot moves sealed reagent vials from storage locations in the cassette to addition positions and also dynamically provides vacuum and inert gas to ports on the cassette. The software integrates these automated features into chemistry unit operations (e.g., React, Evaporate, Add) to intuitively create synthesis protocols. 2-Deoxy-2-[18F]fluoro-5-methyl-β-l-arabinofuranosyluracil (l-[18F]FMAU) and 2-deoxy-2-[18F]fluoro-β-d-arabinofuranosylcytosine (d-[18F]FAC) were synthesized to validate the system. Results l-[18F]FMAU and d-[18F]FAC were successfully synthesized in 165 and 170 min, respectively, with decay-corrected radiochemical yields of 46% ± 1% (n = 6

  19. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    SciTech Connect

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  20. Rapid pressure swing absorption cleanup of post-shift reactor synthesis gas

    SciTech Connect

    Sirkar, K.K.; Majumdar, S.; Bhaumik, S.

    1992-09-24

    This investigation is concerned with the separation of gas mixtures using a novel concept of rapid pressure swing absorption (RAPSAB) of gas in a stationary absorbent liquid through gas-liquid interfaces immobilized in the pore mouths of hydrophobic microporous membranes. The process is implemented in a module well-packed with hydrophobic microporous hollow fiber membranes. Before we proceed to RAPSAB studies with reactive absorbents, it is necessary to make an effort to compare experimental results with those predicted by the models. The only model developed so far involved a type of RAPSAB cycle (Mode 3) for which limited data were acquired earlier. A number of experiments have, therefore, been conducted in this mode to characterize the absorption part of the cycle. A new and more compact module (Module No. 5) was made using 840 fibers and a teflon casing inside the stainless steel shell to exactly define the fiber packing density. This allows an exact calculation of Happel's free surface radius. Experiments were carried out using a CO[sub 2]-N[sub 2] mixture of around 10% CO[sub 2] and balanced N[sub 2] using both modules 4 and 5 over a wide range of absorption times.

  1. Defect formation in aqueous environment: Theoretical assessment of boron incorporation in nickel ferrite under conditions of an operating pressurized-water nuclear reactor (PWR)

    NASA Astrophysics Data System (ADS)

    Rák, Zs.; Bucholz, E. W.; Brenner, D. W.

    2015-06-01

    A serious concern in the safety and economy of a pressurized water nuclear reactor is related to the accumulation of boron inside the metal oxide (mostly NiFe2O4 spinel) deposits on the upper regions of the fuel rods. Boron, being a potent neutron absorber, can alter the neutron flux causing anomalous shifts and fluctuations in the power output of the reactor core. This phenomenon reduces the operational flexibility of the plant and may force the down-rating of the reactor. In this work an innovative approach is used to combine first-principles calculations with thermodynamic data to evaluate the possibility of B incorporation into the crystal structure of NiFe2O4 , under conditions typical to operating nuclear pressurized water nuclear reactors. Analyses of temperature and pH dependence of the defect formation energies indicate that B can accumulate in NiFe2O4 as an interstitial impurity and may therefore be a major contributor to the anomalous axial power shift observed in nuclear reactors. This computational approach is quite general and applicable to a large variety of solids in equilibrium with aqueous solutions.

  2. A model for calculation of RCS pressure during reflux boiling under reduced inventory conditions and its assessment against PKL data. [Reactor Cooling Systems (RCS)

    SciTech Connect

    Palmrose, D.E. ); Mandl, R.M. )

    1991-01-01

    There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.

  3. Distribution of sulfur during coal pyrolysis in a high pressure entrained-flow reactor

    SciTech Connect

    Fatemi-Badi, M.; Scaroni, A.W.; Jenkins, R.G. )

    1988-06-01

    An understanding of the fundamental and interrelated processes involved in coal pyrolysis/gasification is essential in order to advance gasification technologies. Therefore, it is necessary to generate data on the effect of coal properties and operating conditions on coal devolatilization behavior under conditions similar to those in advanced-concept gasifiers; typically, a high-temperature and high pressure environment for entrained coal particles. The thermal decomposition of raw coal produces solid char/coke plus some liquid (tar) and gaseous products. Desulfurization of coal prior to combustion can be achieved by one or more of the following methods. 1) Advanced physical cleaning -- this method involves a variety of approaches (i.e. froth flotation, float-sink, etc.) for reduction of pyrite in the coal based on the differences between the density of pyrite (5.0 g/cm/sup 3/) or marcasite (4.87 g/cm/sup 3/) and the organic matrix (1.2 - 1.5 g/cm/sup 3/). 2) Chemical coal cleaning -- based on the concept of breaking the chemical bonds of the organic sulfur by exposing the coal usually to motten chemicals such as sodium hydroxide. 3) Conversion of the coal to low-sulfur liquid and gaseous fuels -- based on liquefaction and gasification technologies. Pyrolysis has been used by some researchers to study the behavior and distribution of sulfur in coal. Some investigators have also used pyrolysis in order to study the chemistry and kinetics of reactions involving sulfur-containing compounds in coal.

  4. APFIM characterization of a high phosphorus Russian RPV weld

    NASA Astrophysics Data System (ADS)

    Miller, M. K.; Russell, K. F.

    1996-03-01

    A microstructural characterization of a high phosphorus (0.035 wt% P) weld from the pressure vessel of a Russian VVER nuclear reactor has been performed. The microstructure of these steels consists of intragranular and intergranular vanadium carbonitride precipitates of average composition 51.3 ± 0.9 at% V, 18.8 ± 0.7 at% C, 22.1 ± 0.7 at% N, 4.9 ± 0.4 at% Cr, 2.4 ± 0.3 at% Mo, 0.36 ± 0.05 at% Fe, 0.07 ± 0.05 at% B and 0.03 ± 0.03 at% P. The lath and grain boundaries were also coated with a thin film of molybdenum carbonitride precipitates. The phosphorus coverage at the boundaries in the unirradiated material was ˜ 13% of a monolayer in agreement with predictions from the McLean model of equilibrium segregation. After neutron irradiation to a fluence of 1.15 × 10 20 n cm -2, the phosphorus coverage had increased significantly to up to ˜ 60% of a monolayer. This result indicates that neutron irradiation significantly enhanced the phosphorus segregation process. Phosphorus and copper clusters were also observed in the matrix of the neutron-irradiated material.

  5. Determining the optimal transmembrane gas pressure for nitrification in membrane-aerated biofilm reactors based on oxygen profile analysis.

    PubMed

    Wang, Rongchang; Xiao, Fan; Wang, Yanan; Lewandowski, Zbigniew

    2016-09-01

    The goal of this study was to investigate the effect of transmembrane gas pressure (P g) on the specific ammonium removal rate in a membrane-aerated biofilm reactor (MABR). Our experimental results show that the specific ammonium removal rate increased from 4.98 to 9.26 gN m(-2) day(-1) when P g increased from 2 to 20 kPa in an MABR with a biofilm thickness of approximately 600 μm. However, this improvement was not linear; there was a threshold of P g separating the stronger and weaker effects of P g. The ammonium removal rate was improved less significantly when P g was over the threshold, indicating that there is an optimal threshold of P g for maximizing ammonium removal in an MABR. The change in oxygen penetration depth (d p) is less sensitive to P g in the ammonia-oxidizing active layer than in the inactive layer in membrane-aerated biofilm. The location of the P g threshold is at the same point as the thickness of the active layer on the curve of d p versus P g; thus, the active layer thickness and the optimal P g can be determined on the basis of the changes in the slope of d p to P g. PMID:27170321

  6. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    SciTech Connect

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  7. Characterisation of interfacial segregation to Cu-enriched precipitates in two thermally aged reactor pressure vessel steel welds.

    PubMed

    Styman, P D; Hyde, J M; Wilford, K; Parfitt, D; Riddle, N; Smith, G D W

    2015-12-01

    To understand the contribution of long term thermal ageing to Reactor Pressure Vessel (RPV) embrittlement two high Cu steel welds with different Ni contents were thermally aged for times up to 100,000 h at 330 °C and 365 °C. Microstructural characterisation using Atom Probe Tomography was performed. Thermal ageing produced a high number density of nano-scale Cu-enriched precipitates. The precipitate-matrix interfaces were enriched in Ni, Mn and Si. The characterisation of these interfaces using a double cluster search approach is the subject of this work. The interface region around thermally-induced precipitates was found to be wider in steels with higher bulk Ni contents and where precipitates had larger core radii. The effect of ageing temperature on interface width was small when comparing precipitates of equal core radius. The narrower interface width in the lower Ni steels is reflected in the composition of the interface, which has a lower Ni content than in the higher Ni material. The reduction in interfacial energy due to the segregation of Ni, Mn and Si has been calculated and shows enhanced reductions in interfacial energy with increasing precipitate size, but no obvious effect of temperature. PMID:26051655

  8. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    PubMed

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. PMID:27209090

  9. The role of intergranular chromium carbides on intergranular oxidation of nickel based alloys in pressurized water reactors primary water

    NASA Astrophysics Data System (ADS)

    Gaslain, F. O. M.; Le, H. T.; Duhamel, C.; Guerre, C.; Laghoutaris, P.

    2016-02-01

    Alloy 600 is used in pressurized water reactors (PWRs) but is susceptible to primary water stress corrosion cracking (PWSCC). Intergranular chromium carbides have been found beneficial to reduce PWSCC. Focussed ion beam coupled with scanning electron microscopy (FIB/SEM) 3D tomography has been used to reconstruct the morphology of grain boundary oxide penetrations and their interaction with intergranular Cr carbides in Alloy 600 subjected to a PWR environment. In presence of intergranular Cr carbides, the intergranular oxide penetrations are less deep but larger than without carbide. However, the intergranular oxide volumes normalized by the grain boundary length for both samples are similar, which suggest that intergranular oxidation growth rate is not affected by carbides. Analytical transmission electron microscopy (TEM) shows that the intergranular oxide consists mainly in a spinel-type oxide containing nickel and chromium, except in the vicinity of Cr carbides where Cr2O3 was evidenced. The formation of chromium oxide may explain the lower intergranular oxide depth observed in grain boundaries containing Cr carbides.

  10. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  11. Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.; Suzuki, M.

    2014-09-01

    The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400 °C for periods of 100-10,000 h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100 h, and gradually for aging durations greater than 1000 h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000 h, had the highest number density after aging for 2000 h, and coarsened after aging for 10,000 h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000 h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition.

  12. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Matsukawa, Y.; Nishiyama, Y.; Onizawa, K.

    2012-06-01

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the δ-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

  13. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  14. Effect of neutron irradiation on tensile properties of materials for pressure vessel internals of WWER type reactors

    NASA Astrophysics Data System (ADS)

    Sorokin, A. A.; Margolin, B. Z.; Kursevich, I. P.; Minkin, A. J.; Neustroev, V. S.

    2014-01-01

    Tensile properties of austenitic stainless steels used for pressure vessel internals of WWER type reactors (18Cr-10Ni-Ti steel and its weld metal) in the initial and irradiated conditions were investigated. Based on the presented original investigations and generalization of the available experimental data the dependences of yield strength and ultimate strength on a neutron damage dose up to 108 dpa, irradiation temperature range 320-450 °C and test temperature range 20-450 °C were obtained. The method of determination of the stress-strain curve parameters was proposed which does not require uniform elongation of a specimen as an input parameter. The dependences was proposed allowing one to calculate the stress-strain curve parameters for 18Cr-10Ni-Ti steel and its weld metal for different test temperatures, different irradiation temperatures and doses. The dependences were obtained to describe the fracture strain decrease under irradiation at a temperature range 320-340 °C when irradiation swelling is absent.

  15. Steam Gasification Rates of Three Bituminous Coal Chars in an Entrained-Flow Reactor at Pressurized Conditions

    SciTech Connect

    Lewis, Aaron D.; Holland, Troy M.; Marchant, Nathaniel R.; Fletcher, Emmett G.; Henley, Daniel J.; Fuller, Eric G.; Fletcher, Thomas H.

    2015-02-26

    Three bituminous coal chars (Illinois #6, Utah Skyline, and Pittsburgh #8) were gasified separately at total pressures of 10 and 15 atm in an entrained-flow reactor using gas temperatures up to 1830 K and particle residence times <240 ms. The experiments were performed at conditions where the majority of particle mass release was due to H2O gasification, although select experiments were performed at conditions where significant mass release was due to gasification by both H2O and CO2. The measured coal data we recorded were fit to three char gasification models including a simple first-order global model, as well as the CCKNand CCK models that stem from the CBK model. The optimal kinetic parameters for each of the three models are reported, and the steam reactivity of the coal chars at the studied conditions is as follows: Pittsburgh #8 > Utah Skyline > Illinois #6.

  16. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  17. Russian Lunar Space Program

    NASA Astrophysics Data System (ADS)

    Zelenyi, Lev; Petrukovich, Anatoly; Khartov, Victor V.; Dolgopolov, Vladimir; Mitrofanov, Igor; Martunov, M.; Lukianchikov, A.; Shevchenko, Vladislav

    Russia had a great number of “firsts” in Lunar Studies (first soft landing, first pictures of the dark side of the moon, first sample return, first rover). Now after a long break the focus of Russian Space Program is again aimed to the lunar science investigations. These investigations have two aims: 1) to get answers to a principal questions of lunar formation and evolution, search for volatiles and regions with subsurface lunar permafrost, studies of lunar dust, electrostatic fields and magnetic anomalies. 2) Preparation to Lunar Exploration stage and search for most promising sites for future lunar habitable scientific stations. First stage of Russian Lunar program during this decade of 2 Lunar includes launches Landers and one Lunar orbiter, discussed in a preceding talks. Further steps during the next decade are related, first of all, with the cryogenic lunar sample return from a certain locations, hear South (or North ) poles, which according to the analysis of orbital observations are enriched by the subsurface water ice inclusions. Next steps, which are planned now are transitional to the exploration stage: delivery of a “ heavy rover“ to the specific site (thoroughly investigated during previous stages), accomplishment of technological experiments on the mitigation of lunar dust and space radiation hazards, simple initial experiments on radioastronomy and cosmic ray studies. It is a long and complicated path to go and quite naturally Russia considers that all important steps on this way will be done in international partnership.

  18. Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

    NASA Astrophysics Data System (ADS)

    Ohtomo, Y.; Ijiri, A.; Ikegawa, Y.; Tsutsumi, M.; Imachi, H.; Uramoto, G.; Hoshino, T.; Morono, Y.; Tanikawa, W.; Hirose, T.; Inagaki, F.

    2013-12-01

    The geological CO2 sequestration into subsurface unmineable oil/gas fields and coal formations has been considered as one of the possible ways to reduce dispersal of anthropogenic greenhouse gasses into the atmosphere. However, feasibility of CO2 injection largely depends on a variety of geological and economical settings, and its ecological consequences have remained largely unpredictable. To address these issues, we developed a new flow-through-type CO2 injection system designated as the 'geobio-reactor system' to examine possible geophysical, geochemical and microbiological impact caused by CO2 injection under in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. In this study, we investigated Eocene bituminous coal-sandstones in the northwestern Pacific coast, Hokkaido, Japan, using the geobio-reactor system. Anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 mL/min, respectively) were continuously supplemented into the coal-sand column under the pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. Molecular analysis of bacterial 16S rRNA genes showed that predominant bacterial components were physically dispersed from coal to sand as the intact form during experiment. Cultivation experiments from sub-sampling fluids indicated that some terrestrial microbes could preserve their survival in subsurface condition. Molecular analysis of archaeal 16S rRNA genes also showed that no methanogens were activated during experiment. We also anaerobically incubated the coal sample using conventional batch-type cultivation technique with a medium for methanogens. After one year of the batch incubation at 20°C, methane could be detected from the cultures except for the acetate-fed culture. The sequence of archaeal 16S rRNA genes via PCR amplification obtained from the H2 plus formate-fed culture was affiliated with a hydrogenotrophic methanogen within the genus Methanobacterium, whereas the methanol plus trimethylamine culture

  19. Pressure regulator

    DOEpatents

    Ebeling, Jr., Robert W.; Weaver, Robert B.

    1979-01-01

    The pressure within a pressurized flow reactor operated under harsh environmental conditions is controlled by establishing and maintaining a fluidized bed of uniformly sized granular material of selected density by passing the gas from the reactor upwardly therethrough at a rate sufficient to fluidize the bed and varying the height of the bed by adding granular material thereto or removing granular material therefrom to adjust the backpressure on the flow reactor.

  20. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    DOE PAGESBeta

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-04-20

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3 and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistrymore » in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from VOCs, NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach two orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s−1) ten-fold OH suppression is observed at OHRext ∼ 100 s−1, which is similar or lower than used in many laboratory studies. This finding may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over an order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross-sections in the model are within ± 25% for OH exposure and within ± 60% for other parameters. These uncertainties are small relative to the dynamic range of outputs. Uncertainty

  1. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    NASA Astrophysics Data System (ADS)

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-04-01

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3 and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistry in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from VOCs, NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach two orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s-1) ten-fold OH suppression is observed at OHRext ∼ 100 s-1, which is similar or lower than used in many laboratory studies. This finding may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over an order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross-sections in the model are within ± 25% for OH exposure and within ± 60% for other parameters. These uncertainties are small relative to the dynamic range of outputs. Uncertainty analysis

  2. Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

    SciTech Connect

    Nad, Shreya; Gu, Yajun; Asmussen, Jes

    2015-07-15

    The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.

  3. Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions.

    PubMed

    Nad, Shreya; Gu, Yajun; Asmussen, Jes

    2015-07-01

    The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100-260 Torr pressure range and 1.5-2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (η(coup)) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance. PMID:26233399

  4. Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

    NASA Astrophysics Data System (ADS)

    Nad, Shreya; Gu, Yajun; Asmussen, Jes

    2015-07-01

    The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100-260 Torr pressure range and 1.5-2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (ηcoup) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.

  5. Evaluation of cracking in feedwater piping adjacent to the steam generators in Nine Pressurized Water Reactor Plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.; Scott, R.G.

    1980-06-25

    Cracking in ASTM A106-B and A106-C feedwater piping was detected near the inlet to the steam generators in a number of pressurized water reactor plants. We received sections with cracks from nine of the plants with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Variations were observed in piping surface irregularities, corrosion-product, pit, and crack morphology, surface elmental and crystal structure analyses, and steel microstructures and mechanical properties. However, with but two exceptions, namely, arrest bands and major surface irregularities, we were unable to relate the extent of cracking to any of these factors. Tensile and fracture toughness (J/sub Ic/ and tearing modulus) properties were measured over a range of temperatures and strain rates. No unusual properties or microstructures were observed that could be related to the cracking problem. All crack surfaces contained thick oxide deposits and showed evidence of cyclic events in the form of arrest bands. Transmission electron microscopy revealed fatigue striations on replicas of cleaned crack surfaces from one plant and possibly from three others. Calculations based on the observed striation spacings gave a value of ..delta..sigma = 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses. Although surface irregularities and corrosion pits were sources for crack initiation and corrosion may have contributed to crack propagation, it is proposed that the overriding factor in the cracking problem is the presence of unforeseen cyclic loads.

  6. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  7. Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

    NASA Astrophysics Data System (ADS)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

    2016-08-01

    In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

  8. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    SciTech Connect

    Syarifah, Ratna Dewi Suud, Zaki

    2015-09-30

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  9. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  10. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    SciTech Connect

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  11. Russian Orthography and Learning to Read

    ERIC Educational Resources Information Center

    Kerek, Eugenia; Niemi, Pekka

    2009-01-01

    The unique structure of Russian orthography may influence the organization and acquisition of reading skills in Russian. The present review examines phonemic-graphemic correspondences in Russian orthography and discusses its grain-size units and possible difficulties for beginning readers and writers. Russian orthography is governed by a…

  12. Inside the Russian Soyuz Spacecraft

    NASA Video Gallery

    In this episode of NASA: Behind the Scenes, take a tour inside the Russian Soyuz, the vehicle which takes the expedition crews back and forth to the International Space Station. Astronaut Mike Finc...

  13. Russian Aviation and Space Agency

    NASA Astrophysics Data System (ADS)

    Murdin, P.

    2000-11-01

    The Russian Space Agency (RKA) was created on 25 February 1992 by a decree issued by the President of the Russian Federation. It was formed after the break-up of the former Soviet Union and the dissolution of the Soviet space program. The RKA uses the technology and launch sites that belonged to the former Soviet space program. This includes payment to Kazakhstan for use of the Baikonur Cosmodrom...

  14. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    SciTech Connect

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  15. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Edmondson, P. D.; Miller, M. K.; Powers, K. A.; Nanstad, R. K.

    2016-03-01

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m-2 (E > 1 MeV), and inlet temperatures of ∼289 °C (∼552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m-3, this copper level was below the solubility limit. A number density of 2 × 1022 m-3 of Ni-, Mn- Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m-3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m-3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain boundary in the low fluence

  16. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGESBeta

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  17. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    SciTech Connect

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

  18. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    SciTech Connect

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  19. Pressurized pyrolysis of rice husk in an inert gas sweeping fixed-bed reactor with a focus on bio-oil deoxygenation.

    PubMed

    Qian, Yangyang; Zhang, Jie; Wang, Jie

    2014-12-01

    The pyrolysis of rice husk was conducted in a fixed-bed reactor with a sweeping nitrogen gas to investigate the effects of pressure on the pyrolytic behaviors. The release rates of main gases during the pyrolysis, the distributions of four products (char, bio-oil, water and gas), the elemental compositions of char, bio-oil and gas, and the typical compounds in bio-oil were determined. It was found that the elevation of pressure from 0.1MPa to 5.0MPa facilitated the dehydration and decarboxylation of bio-oil, and the bio-oils obtained under the elevated pressures had significantly less oxygen and higher calorific value than those obtained under atmospheric pressure. The former bio-oils embraced more acetic acid, phenols and guaiacols. The elevation of pressure increased the formation of CH4 partially via the gas-phase reactions. An attempt is made in this study to clarify "the pure pressure effect" and "the combined effect with residence time". PMID:25463787

  20. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    DOE PAGESBeta

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Lee-Taylor, J.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-11-20

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3, and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistrymore » in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from volatile organic compounds (VOCs), NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach 2 orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s−1), 10-fold OH suppression is observed at OHRext ~ 100 s−1, which is similar or lower than used in many laboratory studies. The range of modeled OH suppression for literature experiments is consistent with the measured values except for those with isoprene. The finding on OH suppression may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over a 1-order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross

  1. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    NASA Astrophysics Data System (ADS)

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Lee-Taylor, J.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-11-01

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3, and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistry in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from volatile organic compounds (VOCs), NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach 2 orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s-1), 10-fold OH suppression is observed at OHRext ~ 100 s-1, which is similar or lower than used in many laboratory studies. The range of modeled OH suppression for literature experiments is consistent with the measured values except for those with isoprene. The finding on OH suppression may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over a 1-order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross-sections in the

  2. USING LIGA BASED MICROFABRICATION TO IMPROVE OVERALL HEAT TRANSFER EFFICIENCY OF PRESSURIZED WATER REACTOR: I. Effects of Different Micro Pattern on Overall Heat Transfer.

    SciTech Connect

    Zhang, M.; Ibekwe, S.; Li, G.; Pang, S.S.; and Lian, K.

    2006-07-01

    The Pressurized Water Reactors (PWRs in Figure 1) were originally developed for naval propulsion purposes, and then adapted to land-based applications. It has three parts: the reactor coolant system, the steam generator and the condenser. The Steam generator (a yellow area in Figure 1) is a shell and tube heat exchanger with high-pressure primary water passing through the tube side and lower pressure secondary feed water as well as steam passing through the shell side. Therefore, a key issue in increasing the efficiency of heat exchanger is to improve the design of steam generator, which is directly translated into economic benefits. The past research works show that the presence of a pin-fin array in a channel enhances the heat transfer significantly. Hence, using microfabrication techniques, such as LIGA, micro-molding or electroplating, some special microstructures can be fabricated around the tubes in the heat exchanger to increase the heat-exchanging efficiency and reduce the overall size of the heat-exchanger for the given heat transfer rates. In this paper, micro-pin fins of different densities made of SU-8 photoresist are fabricated and studied to evaluate overall heat transfer efficiency. The results show that there is an optimized micro pin-fin configuration that has the best overall heat transfer effects.

  3. A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions

    SciTech Connect

    Morfin, Franck; Piccolo, Laurent

    2013-09-15

    A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup −4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

  4. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  5. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  6. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    SciTech Connect

    Bowman, S.M.; Suto, T. |

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

  7. Issues in the flight qualification of a space power reactor

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-10-01

    This paper presents an overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP). The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between US and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year.

  8. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  9. Russian Prime Minister Calls the Station Crew

    NASA Video Gallery

    Russian Prime Minister Vladimir Putin called the International Space Station from the Russian Mission Control Center in Korolev, Russia, on Jan. 11, 2011. Putin also offered his condolences to ISS ...

  10. Russian Soyuz in Launch Position

    NASA Technical Reports Server (NTRS)

    2000-01-01

    The Soyuz TM-31 launch vehicle is shown in the vertical position for its launch from Baikonur, carrying the first resident crew to the International Space Station. The Russian Soyuz launch vehicle is an expendable spacecraft that evolved out of the original Class A (Sputnik). From the early 1960s until today, the Soyuz launch vehicle has been the backbone of Russia's marned and unmanned space launch fleet. Today, the Soyuz launch vehicle is marketed internationally by a joint Russian/French consortium called STARSEM. As of August 2001, there have been ten Soyuz missions under the STARSEM banner.

  11. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    SciTech Connect

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  12. Non-Textbook Materials in Russian Civilization Classes

    ERIC Educational Resources Information Center

    Vinograde, Ann C.

    1977-01-01

    Supplementary, non-textbook materials used in Russian civilization classes include "Paintings from the Russian Museum Collection," with text in Russian and English; a Russian desk calendar; "In the Russian Style," with costume photographs and poetry excerpts, and a Russian primary school book. Uses of these materials are discussed. (CHK)

  13. A Differential Pressure Laminar Flow Reactor Supports Osteogenic Differentiation and Extracellular Matrix Formation from Adipose Mesenchymal Stem Cells in a Macroporous Ceramic Scaffold

    PubMed Central

    Kasper, Cornelia; Israelowitz, Meir; Gille, Christoph; von Schroeder, Herbert P.; Reimers, Kerstin; Vogt, Peter M.

    2012-01-01

    Abstract We present a laminar flow reactor for bone tissue engineering that was developed based on a computational fluid dynamics model. The bioreactor design permits a laminar flow field through its specific internal shape. An integrated bypass system that prevents pressure build-up through bypass openings for pressure release allows for a constant pressure environment during the changing of permeability values that are caused by cellular growth within a porous scaffold. A macroporous ceramic scaffold, composed of zirconium dioxide, was used as a test biomaterial that studies adipose stem cell behavior within a controlled three-dimensional (3D) flow and pressure environment. The topographic structure of the material provided a basis for stem cell proliferation and differentiation toward the osteogenic lineage. Dynamic culture conditions in the bioreactor supported cell viability during long-term culture and induced cell cluster formation and extra-cellular matrix deposition within the porous scaffold, though no complete closure of the pores with new-formed tissue was observed. We postulate that our system is suitable for studying fluid shear stress effects on stem cell proliferation and differentiation toward bone formation in tissue-engineered 3D constructs. PMID:23515420

  14. The TACIS Nuclear Programme: Assistance in Upgrading Russian Nuclear Power Stations - An Overview of the Individual Projects in the Internet

    SciTech Connect

    Bieth, Michel; Schoels, Hubert

    2006-07-01

    The European Union' TACIS1 programme has been established for the New Independent States (NIS), among them in the Russian Federation since 1991. One priority of TACIS funding is Nuclear Safety. The European Commission has made available a total of 944 Million Euros for nuclear safety programmes covering the period 1991-2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. JRC is carrying out works in the following areas: On-Site Assistance for TACIS operating Nuclear Power Plants; Design Safety and Dissemination of TACIS results; Reactor Pressure Vessel Embrittlement for VVER; Regulatory Assistance; Industrial Waste Management; Nuclear Safeguards; All TACIS projects, dealing with these areas of activity are now available in so called Project Description Sheets (PDS) or Project Results Sheets (PRS) in the Internet for everybody. JRC has created in the Internet an easy to open and to browse database which contains the result of works in relation to the above mentioned nuclear activities. This presentation gives an on-line overview of the app. 430 projects which have been implemented so far since the outset of the TACIS Nuclear Programme in the Russian Federation, which is representative to the other CIS countries, benefiting from the TACIS. The presentation will mainly consist of an on-line-demonstration of the TACIS Nuclear WEB Page, created by JRC. (authors)

  15. Teaching Russian Culture: Concepts of Nature.

    ERIC Educational Resources Information Center

    Jones, Dianne

    Russian culture is very much influenced by its huge land area, peculiar topography, and harsh climate. To understand Russian culture one must know how Russians perceive nature. This paper discusses how this concept may be conveyed to U.S. middle school students through poetry. Poems about nature can provide students an opportunity to understand…

  16. Processing of Contrastiveness by Heritage Russian Bilinguals

    ERIC Educational Resources Information Center

    Sekerina, Irina A.; Trueswell, John C.

    2011-01-01

    Two eye-tracking experiments in the Visual World paradigm compared how monolingual Russian (Experiment 1) and heritage Russian-English bilingual (Experiment 2) listeners process contrastiveness online in Russian. Materials were color adjective-noun phrases embedded into the split-constituent construction Krasnuju polozite zvezdovku..."Red put…

  17. Russian Loanword Adaptation in Persian; Optimal Approach

    ERIC Educational Resources Information Center

    Kambuziya, Aliye Kord Zafaranlu; Hashemi, Eftekhar Sadat

    2011-01-01

    In this paper we analyzed some of the phonological rules of Russian loanword adaptation in Persian, on the view of Optimal Theory (OT) (Prince & Smolensky, 1993/2004). It is the first study of phonological process on Russian loanwords adaptation in Persian. By gathering about 50 current Russian loanwords, we selected some of them to analyze. We…

  18. An Evolutionary Sketch of Russian Kinship.

    ERIC Educational Resources Information Center

    Friedrich, Paul

    This paper presents a preliminary sketch of the evolution of Russian kinship from the reconstructed stages of Proto-Indo-European and Proto-Slavic, through old and nineteenth century Russian, to the trends of contemporary modern Russian. Linguistic, historical, and anthropological approaches have been combined. The kinship terminology is viewed as…

  19. Social and Racial Correlates of Russian Roulette

    ERIC Educational Resources Information Center

    Stack, Steven; Wasserman, Ira

    2008-01-01

    The epidemiology of a neglected form of suicidal behavior, Russian roulette, is addressed. Also tested is an explanation of racial differences based on the opportunity theory of deviant behavior related to the availability of revolvers, necessary weapons with which to play Russian roulette. Data refer to 15 cases of Russian roulette found through…

  20. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    SciTech Connect

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M.

    2013-09-11

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ∼5% of maximum value by the final moment of the stage of linear increase in the temperature.

  1. Simulation of the spatial distribution of the acoustic pressure in sonochemical reactors with numerical methods: a review.

    PubMed

    Tudela, Ignacio; Sáez, Verónica; Esclapez, María Deseada; Díez-García, María Isabel; Bonete, Pedro; González-García, José

    2014-05-01

    Numerical methods for the calculation of the acoustic field inside sonoreactors have rapidly emerged in the last 15 years. This paper summarizes some of the most important works on this topic presented in the past, along with the diverse numerical works that have been published since then, reviewing the state of the art from a qualitative point of view. In this sense, we illustrate and discuss some of the models recently developed by the scientific community to deal with some of the complex events that take place in a sonochemical reactor such as the vibration of the reactor walls and the nonlinear phenomena inherent to the presence of ultrasonic cavitation. In addition, we point out some of the upcoming challenges that must be addressed in order to develop a reliable tool for the proper designing of efficient sonoreactors and the scale-up of sonochemical processes. PMID:24355287

  2. Morphological Rules in Russian Conjugation.

    ERIC Educational Resources Information Center

    Thomas, Linda Kopp

    Recent analyses of Russian (Halle 1963, Lightner 1972) have been forced by the criteria of rule "naturalness" and rule "generality" to posit highly abstract underlying forms. These underlying forms and rules are claimed to represent the speaker's competence. Such analyses are now being criticized (Derwing 1973, Hooper 1974) on the following…

  3. Oral Proficiency Testing in Russian.

    ERIC Educational Resources Information Center

    Wing, Barbara H., Comp.; Mayewski, Sandi, Comp.

    A handbook compiled for use in a one-day workshop on oral proficiency testing for teachers of Russian gives an overview of oral proficiency assessment principles and the available techniques. One section explains the American Council on the Teaching of Foreign Languages/Educational Testing Service (ACTFL/ETS) Oral Proficiency Interview process and…

  4. Russian Honeybee Breeders Association Manual

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The USDA-Agricultural Research Service, Honey Bee Breeding, Genetics and Physiology Laboratory has produced a stock of bees called “Russian Honey Bees”. This name recognizes the original source of the honey bee queens that were used to form the stock. Queen bees from Russia’s far-east, many o...

  5. Nuclear safety research collaborations between the U.S. and Russian Federation International Nuclear Safety Centers

    SciTech Connect

    Hill, D. J.; Braun, J. C.; Klickman, A. E.; Bougaenko, S. E.; Kabonov, L. P.; Kraev, A. G.

    2000-05-05

    The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) have formed International Nuclear Safety Centers to collaborate on nuclear safety research. USDOE established the US Center (ISINSC) at Argonne National Laboratory (ANL) in October 1995. MINATOM established the Russian Center (RINSC) at the Research and Development Institute of Power Engineering (RDIPE) in Moscow in July 1996. In April 1998 the Russian center became a semi-independent, autonomous organization under MINATOM. The goals of the center are to: Cooperate in the development of technologies associated with nuclear safety in nuclear power engineering; Be international centers for the collection of information important for safety and technical improvements in nuclear power engineering; and Maintain a base for fundamental knowledge needed to design nuclear reactors. The strategic approach is being used to accomplish these goals is for the two centers to work together to use the resources and the talents of the scientists associated with the US Center and the Russian Center to do collaborative research to improve the safety of Russian-designed nuclear reactors. The two centers started conducting joint research and development projects in January 1997. Since that time the following ten joint projects have been initiated: INSC databases--web server and computing center; Coupled codes--Neutronic and thermal-hydraulic; Severe accident management for Soviet-designed reactors; Transient management and advanced control; Survey of relevant nuclear safety research facilities in the Russian Federation; Computer code validation for transient analysis of VVER and RBMK reactors; Advanced structural analysis; Development of a nuclear safety research and development plan for MINATOM; Properties and applications of heavy liquid metal coolants; and Material properties measurement and assessment. Currently, there is activity in eight of these projects. Details on each of these

  6. Russians Work on Aft Portion of Zarya

    NASA Technical Reports Server (NTRS)

    1997-01-01

    In this photograph, Russians are working on the aft portion of the United States-funded, Russian-built Functional Cargo Bay (FGB) also known as Zarya (Russian for sunrise). Built at Khrunichev, the FGB began pre-launch testing shortly after this photo was taken. Launched by a Russian Proton rocket from the Baikonu Cosmodrome on November 20, 1998, Zarya was the first element of the International Space Station (ISS) followed by the U.S. Unity Node. The aft docking mechanism, Pirs, on the far right with ventilation ducting rurning through it, will be docked with the third Station element, the Russian Service Module, or Zvezda.

  7. Russian perspectives: The past shapes the present

    SciTech Connect

    Houck, R.P.

    1994-11-01

    This document contains an outline of a speech given to a group of professionals at Pacific Northwest Laboratory which was intended to give an unbiased view of Soviet perceptions. Topics discussed include: The new mission of US and Soviet labs and institutions to develop products and dedicate research to post cold war threat, historical prospectives of Russia, Russian military roles and missions, ideology of Russian politics, evils of capitalism, Russian civil war, communism, world war II, Russian losses during the war, the cold war, reasons why America should care what happens in Russia, the internal threat against a market-based economy, the US should help, and the Russian people and their attitudes.

  8. Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

    2001-02-01

    Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

  9. Accelerating the Reduction of Excess Russian Highly Enriched Uranium

    SciTech Connect

    Benton, J; Wall, D; Parker, E; Rutkowski, E

    2004-02-18

    This paper presents the latest information on one of the Accelerated Highly Enriched Uranium (HEU) Disposition initiatives that resulted from the May 2002 Summit meeting between Presidents George W. Bush and Vladimir V. Putin. These initiatives are meant to strengthen nuclear nonproliferation objectives by accelerating the disposition of nuclear weapons-useable materials. The HEU Transparency Implementation Program (TIP), within the National Nuclear Security Administration (NNSA) is working to implement one of the selected initiatives that would purchase excess Russian HEU (93% 235U) for use as fuel in U.S. research reactors over the next ten years. This will parallel efforts to convert the reactors' fuel core from HEU to low enriched uranium (LEU) material, where feasible. The paper will examine important aspects associated with the U.S. research reactor HEU purchase. In particular: (1) the establishment of specifications for the Russian HEU, and (2) transportation safeguard considerations for moving the HEU from the Mayak Production Facility in Ozersk, Russia, to the Y-12 National Security Complex in Oak Ridge, TN.

  10. Pressure dependent calibration of the OH and HOx channels of a FAGE HOx instrument using the Highly Instrumented Reactor for Atmospheric Chemistry (HIRAC)

    NASA Astrophysics Data System (ADS)

    Winiberg, F. A. F.; Smith, S. C.; Bejan, I.; Brumby, C. A.; Ingham, T.; Malkin, T. L.; Orr, S. C.; Heard, D. E.; Seakins, P. W.

    2014-07-01

    The calibration of field instruments used to measure concentrations of OH and HO2 worldwide have traditionally relied on a single method utilising the photolysis of water vapour in air in a flow tube at atmospheric pressure. Here the calibration of two FAGE (Fluorescence Assay by Gaseous Expansion) apparatuses designed for HOx (OH and HO2) measurements have been investigated as a function of external pressure and temperature, using two different laser systems. The conventional method of generating known concentrations of HOx from H2O vapour photolysis in a turbulent flowtube impinging just outside the FAGE sample inlet has been used to study instrument sensitivity as a function of internal fluorescence cell pressure (1.8-3.8 mbar). An increase in the calibration constants COH and CHO2 with pressure was observed and an empirical linear regression of the data was used to describe the trends, with ΔCOH = (17 ± 11)% and ΔCHO2 = (31.6 ± 4.4)% increase per mbar air (uncertainties quoted to 2σ). Presented here are the first direct measurements of the FAGE calibration constants as a function of external pressure (440-1000 mbar) in a controlled environment using the University of Leeds HIRAC chamber (Highly Instrumented Reactor for Atmospheric Chemistry). Two methods were used: the temporal decay of hydrocarbons for calibration of OH, and the kinetics of the second-order recombination of HO2 for HO2 calibrations. Over comparable conditions for the FAGE cell, the two alternative methods are in good agreement with the conventional method, with the average ratio of calibration factors (conventional : alternative) across the entire pressure range COH(conv)/COH(alt) = 1.19 ± 0.26 and CHO2(conv)/CHO2(alt) = 0.96 ± 0.18 (2σ). These alternative calibration methods currently have comparable systematic uncertainties than the conventional method: ~28% and ~41% for the alternative OH and HO2 calibration methods respectively compared to 35% for the H2O vapour photolysis method

  11. Pressure-dependent calibration of the OH and HO2 channels of a FAGE HOx instrument using the Highly Instrumented Reactor for Atmospheric Chemistry (HIRAC)

    NASA Astrophysics Data System (ADS)

    Winiberg, F. A. F.; Smith, S. C.; Bejan, I.; Brumby, C. A.; Ingham, T.; Malkin, T. L.; Orr, S. C.; Heard, D. E.; Seakins, P. W.

    2015-02-01

    The calibration of field instruments used to measure concentrations of OH and HO2 worldwide has traditionally relied on a single method utilising the photolysis of water vapour in air in a flow tube at atmospheric pressure. Here the calibration of two FAGE (fluorescence assay by gaseous expansion) apparatuses designed for HOx (OH and HO2) measurements have been investigated as a function of external pressure using two different laser systems. The conventional method of generating known concentrations of HOx from H2O vapour photolysis in a turbulent flow tube impinging just outside the FAGE sample inlet has been used to study instrument sensitivity as a function of internal fluorescence cell pressure (1.8-3.8 mbar). An increase in the calibration constants CHO and CHO2 with pressure was observed, and an empirical linear regression of the data was used to describe the trends, with ΔCHO = (17 ± 11) % and ΔCHO2 = (31.6 ± 4.4)% increase per millibar air (uncertainties quoted to 2σ). Presented here are the first direct measurements of the FAGE calibration constants as a function of external pressure (440-1000 mbar) in a controlled environment using the University of Leeds HIRAC chamber (Highly Instrumented Reactor for Atmospheric Chemistry). Two methods were used: the temporal decay of hydrocarbons for calibration of OH, and the kinetics of the second-order recombination of HO2 for HO2 calibrations. Over comparable conditions for the FAGE cell, the two alternative methods are in good agreement with the conventional method, with the average ratio of calibration factors (conventional : alternative) across the entire pressure range, COH(conv)/COH(alt) = 1.19 ± 0.26 and CHO2(conv)/CHO2(alt) = 0.96 ± 0.18 (2σ). These alternative calibration methods currently have comparable systematic uncertainties to the conventional method: ~ 28% and ~ 41% for the alternative OH and HO2 calibration methods respectively compared to 35% for the H2O vapour photolysis method; ways in

  12. J-integral elastic plastic fracture mechanics evaluation of the stability of cracks in nuclear reactor pressure vessels

    SciTech Connect

    Gomez, M. P.; McMeeking, R. M.; Parks, D. M.

    1980-06-01

    Contributions were made toward developing a new methodology to assess the stability of cracks in pressure vessels made from materials that exhibit a significant increase in toughness during the early increments of crack growth. It has a wide range of validity from linear elastic to fully plastic behavior.

  13. A criterion for predicting thermal stratification due to high-pressure injection in a circulating reactor loop

    SciTech Connect

    Nourbakhsh, H.P.; Theofanous, T.G.

    1986-09-01

    A physically based criterion for predicting thermal stratification due to high-pressure injection in the presence of loop flow is presented. The predictions are consistent with the CREARE 1/5 scale, CREARE 1/2 scale, and HDR test results.

  14. The thermal circuit of a nuclear power station's unit built around a supercritical-pressure water-cooled reactor

    NASA Astrophysics Data System (ADS)

    Silin, V. A.; Zorin, V. M.; Tagirov, A. M.; Tregubova, O. I.; Belov, I. V.; Povarov, P. V.

    2010-12-01

    Main results obtained from calculations of the steam generator and thermal circuit of the steam turbine unit for a nuclear power unit with supercritical-pressure water coolant and integral layout are presented. The obtained characteristics point to the advisability of carrying out further developments of this promising nuclear power technology.

  15. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    SciTech Connect

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-05-15

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important.

  16. Modeling and Numerical Investigation of the Process of Vapor-Oxygen Gasification of Solid Fuels in a Vertical Flow Reactor Under Pressure

    NASA Astrophysics Data System (ADS)

    Rokhman, B. B.

    2014-09-01

    With the use of the developed model, detailed information has been obtained on the working process in a flow reactor with single- and two-stage schemes of vapor-oxygen gasification of coals under a pressure of 3 MPa. The dependence of the ratios of mass flow rates O2/coal and H2O/coal on the type of fuel has been established and their optimal values for the "Shell" process have been found. At a given consumption ratio of gas coal and brown coal of brand B1, the optimum diameters of particles providing minimum combustible loss of the carbon mixture have been determined. It has been found that the content of methane in the syngas in the case of two-stage gasification is much higher than in the case of single-stage gasification.

  17. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    NASA Astrophysics Data System (ADS)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  18. Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data

    SciTech Connect

    McAfee, W.J.; Bass, B.R.; Bryson, J.W.

    1998-07-01

    A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

  19. Effects of alloying elements on radiation hardening based on loop formation of electron-irradiated light water reactor pressure vessel model steels

    NASA Astrophysics Data System (ADS)

    Nishi, Takakuni; Hashimoto, N.; Ohnuki, S.; Yamamoto, T.; Odette, G. R.

    2011-10-01

    Electron irradiations using a high voltage electron microscope were conducted on several reactor pressure vessel model alloys in order to investigate the effects of alloying elements on the formation and development of defect clusters. In addition, the effects of alloying elements on yield stress change after irradiation were considered, comparing the mean size and number density of dislocation loops with the irradiation-induced hardening. High Cu alloys formed Cu and Mn-Ni-Si rich clusters, and these are important in determining the yield stress increase. High Ni alloys formed a high density of small dislocation loops and probably Mn-Ni-Si rich cluster, which have the effect of increasing the yield stress. High P enhanced radiation-induced segregation on grain boundary, helping prevent dislocation movement.

  20. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.