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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. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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)

  15. Summary of near-term options for Russian plutonium production reactors

    SciTech Connect

    Newman, D.F.; Gesh, C.J.; Love, E.F.; Harms, S.L.

    1994-07-01

    The Russian Federation desires to phase out the production of weapons-grade plutonium. To this end, ten graphite-moderated, water-cooled reactors have been shut down during the last several years. However, complete cessation of plutonium production is impeded because the three operating Russian reactors supply district heat and electricity to the Tomsk and Krasnoyarsk regions in addition to producing weapon-grade plutonium. In August 1992 the Russian Federation Ministry of Atomic Energy (MINATOM) and the Russian Nuclear Regulatory Agency (GAN) requested U.S. assistance for achieving a cessation of weapons-grade plutonium production, placing the plutonium production reactors under safeguards, and conducting a program to evaluate and assist in the upgrade of plant safety. As a result of that and subsequent communications, Secretary O`Leary and Minister Mikhailov have signed a protocol that expressed their desire to shut down the three remaining plutonium production reactors as soon as possible by replacing them with alternate energy sources. In the meantime, both MINATOM and the Department of Energy (DOE) are concerned about the safety of the plants as well as the difficulty in ceasing the production of plutonium as long as the plants continue to operate. A military subsidy has been provided for operation of the production reactor complex. Revenues received for providing district heat and electricity are insufficient to cover costs for the current natural uranium metal fuel cycle. A more economical fuel cycle is needed for civilian operations.

  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. 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.

  18. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  19. Options for monitoring the US Russian bilateral cutoff agreement on shutdown of plutonium production reactors

    SciTech Connect

    Sanborn, J.; Fishbone, L.G.; Lu, Minh-Shih; Stanbro, W.; Libby, R.

    1994-07-01

    Six options are presented for monitoring operating Russian reactors and reprocessing plants under the bilateral cutoff agreement. In order of increasing intrusiveness they are: (A) monitoring of product (oxide or metal) storage only, supplemented with transparency measures at the reactors, (B) monitoring of product storage and reactor operating parameters, to assess reactor plutonium production, (C) monitoring of product storage, reactor operating parameters, and the input accountability tank of the reprocessing plant, (D) monitoring of product storage, the input accountability tank of the reprocessing plant, and application of surveillance to spent fuel, (E) IAEA/NPT-based material accountancy verification without major facility upgrades, and (F) IAEA/NPT-based safeguards, attempting to fulfill IAEA standards for material accountancy accuracy. Each of these options is considered in terms of cost, inspection effort, and effectiveness; however, the paper emphasizes the many uncertainties attendant on such assessments based on our current state of knowledge of these facilities.

  20. 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.

  1. Pressure vessel calculations for VVER-440 reactors.

    PubMed

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  2. 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)

  3. Radiation effects on reactor pressure vessel supports

    SciTech Connect

    Johnson, R.E.; Lipinski, R.E.

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Reactor pressure vessel structural integrity research

    SciTech Connect

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

    1994-12-31

    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 shallow 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 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Using fast reactor component evaluation to pressurized water reactor life extension.

    SciTech Connect

    Allen, T. R.; Cole, J. I.; Kenik, E. A.; Tsai, H.; Ukai, S.; Mizuta, S.; Yoshitake, T.

    1999-10-01

    An understanding of the effects of long-term, low-dose-rate radiation on core components is critical to light-water reactor plant life extension. Following reactor shutdown, materials that had experienced long exposures to low-dose-rate irradiation were retrieved from the EBR-II research reactor for analysis. These components are being analyzed to provide insight into pressurized water reactor life extension. In this work, three examples of EBR-II materials analyses are highlighted: radiation-induced segregation in 304 stainless steel, stress relaxation in Inconel X750, and swelling in 316 stainless steel.;EVALUATION;FAST REACTORS;PWR TYPE REACTORS;

  16. Neutron shielding panels for reactor pressure vessels

    DOEpatents

    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.

  17. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

    2010-07-01

    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

  18. 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.

  19. 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.

  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. 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.

  2. Using fast reactor component evaluation for pressurized water reactor life extension

    NASA Astrophysics Data System (ADS)

    Allen, T. R.; Cole, J. I.; Kenik, E. A.; Tsai, H.; Ukai, S.; Mizuta, S.; Yoshitake, T.

    1999-10-01

    An understanding of the effects of long-term, low-dose-rate radiation on core components is critical to light-water reactor plant life extension. Following reactor shutdown, materials that had experienced long exposures to low-dose-rate irradiation were retrieved from the EBR-II research reactor for analysis. These components are being analyzed to provide insight into pressurized water reactor life extension. In this work, three examples of EBR-II materials analyses are highlighted: radiation-induced segregation in 304 stainless steel, stress relaxation in Inconel X750, and swelling in 316 stainless steel.

  3. 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.

  4. 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.

  5. 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.

  6. An Inexpensive Sampleable Reactor for High-Pressure Chemistry.

    ERIC Educational Resources Information Center

    Shumate, R. E.; Riley, D. P.

    1984-01-01

    Describes how to modify the commercially available and inexpensive Griffin-Worden (Kontes) glass pressure reactors to permit sampling of homogeneous reactions, without changing the relative concentration of species present. All required parts necessary for the modification are commercially available. (JN)

  7. 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

  8. 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.

  9. Neutron flux reduction programs for reactor pressure vessel

    SciTech Connect

    Yoo, C.S.; Kim, B.C.

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  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. 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.

  12. 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.

  13. Upper internals arrangement for a pressurized water reactor

    SciTech Connect

    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.

  14. Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

    NASA Astrophysics Data System (ADS)

    Ivanov, Yu. A.

    2007-12-01

    An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Dual shell pressure balanced reactor vessel. Final project report

    SciTech Connect

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy`s Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R&D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993).

  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. 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

  2. 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.

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

    DOEpatents

    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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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%).

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

  2. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

  3. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

  4. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

  5. REACTOR SERVICES BUILDING, TRA635. FIRST FLOOR PLAN. MOCKUP AND PRESSURE ...

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

    REACTOR SERVICES BUILDING, TRA-635. FIRST FLOOR PLAN. MOCK-UP AND PRESSURE TEST AREAS. ISSUE ROOM, LAUNDRY, STORAGE. IDO MTR-635-IDO-5-A, 6/1953. INL INDEX NO. 531-0635-00-396-110588, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  6. 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

  7. Transient Temperature and Pressure in the Reactor Room During a Core Meltdown Accident

    SciTech Connect

    Shadday, M.A.

    2001-07-17

    The purpose of this numerical model is to determine the optimum ventilation exhaust flow rate for the reactor room. The influence of steam produced in the reactor vessel, on the reactor room pressures, is included in the model. A parametric study of the affect of various steam mass flow rates is included in this document. The affect of steam on the conditions in the reactor room is significant at modest flow rates.

  8. Dual shell reactor vessel: A pressure-balanced system for high pressure and temperature reactions

    SciTech Connect

    Robertus, R.J.; Fassbender, A.G.; Deverman, G.S.

    1995-03-01

    The main purpose of this work was to demonstrate the Dual Shell Pressure Balanced Vessel (DSPBV) as a safe and economical reactor for the hydrothermal water oxidation of hazardous wastes. Experimental tests proved that the pressure balancing piston and the leak detection concept designed for this project will work. The DSPBV was sized to process 10 gal/hr of hazardous waste at up to 399{degree}C (750{degree}F) and 5000 psia (34.5 MPa) with a residence time of 10 min. The first prototype reactor is a certified ASME pressure vessel. It was purchased by Innotek Corporation (licensee) and shipped to Pacific Northwest Laboratory for testing. Supporting equipment and instrumentation were, to a large extent, transported here from Battelle Columbus Division. A special air feed system and liquid pump were purchased to complete the package. The entire integrated demonstration system was assembled at PNL. During the activities conducted for this report, the leak detector design was tested on bench top equipment. Response to low levels of water in oil was considered adequate to ensure safety of the pressure vessel. Shakedown tests with water only were completed to prove the system could operate at 350{degree}C at pressures up to 3300 psia. Two demonstration tests with industrial waste streams were conducted, which showed that the DSPBV could be used for hydrothermal oxidation. In the first test with a metal plating waste, chemical oxygen demand, total organic carbon, and cyanide concentrations were reduced over 90%. In the second test with a munitions waste, the organics were reduced over 90% using H{sub 2}O{sub 2} as the oxidant.

  9. 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.

  10. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  11. 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.

  12. 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.

  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. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Standard- and extended-burnup PWR (pressurized-water reactor) and BWR (boiling-water reactor) reactor models for the ORIGEN2 computer code

    SciTech Connect

    Ludwig, S.B.; Renier, J.P.

    1989-12-01

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs.

  20. 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.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

    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.

  3. The design, construction and three dimensional modeling of a high pressure organometallic chemical vapor deposition reactor

    NASA Astrophysics Data System (ADS)

    McCall, Sonya Denise

    Two high pressure reactors have been designed, built and tested, in order to extend Organometallic Chemical Vapor Deposition (OMCVD) to materials that exhibit large thermal decomposition pressures at their optimum growth temperature. The Differentially Pressure Controlled (DPC) Reactor System was designed and built for use at pressures ≤10 atm. A second generation reactor, the Compact Hard Shell (CHS) Reactor was built in order to extend pressures ≤100 atm. A physico-chemical model of the High Pressure Organometallic Chemical Vapor Deposition (HPOMCVD) process that describes three dimensional transport phenomena as well as gas-phase and surface reactions underlying the growth of compound semiconductors is presented. A reduced-order model of the Organometallic Chemical Vapor Deposition of InN from trimethylindium and ammonia at elevated pressures has been developed and tested. The model describes the flow dynamics coupled to chemical reactions and transport in the flow channel of the Compact Hard Shell Reactor, as a function of substrate temperature, total pressure and centerline flow velocity.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. 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

  9. 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.

  10. 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.

  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. 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…

  13. 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

  14. 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.

  15. 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.

  16. Pressure Vessel and Internals of the International Reactor Innovative and Secure

    SciTech Connect

    Lombardi, C.V.; Padovani, E.; Cammi, A.; Collado, J.M.; Santoro, R.T.; Barnes, J.M.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel. (authors)

  17. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Jumel, Stéphanie; Van-Duysen, Jean Claude

    2005-04-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called ';Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, …) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program.

  18. 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

  19. 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.

  20. 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.

  1. Pressure suppression containment system for boiling water reactor

    DOEpatents

    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.

  2. 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.

  3. Numerical Simulations of Low Pressure Inductively Coupled Plasmas in Geometrically Complex Reactors

    NASA Astrophysics Data System (ADS)

    Yu, Ben; Wu, Hanming; Krishnan, Anantha

    1996-10-01

    A two-dimensional fluid model has been developed for simulation of low pressure inductively coupled plasma (ICP) reactors. The model obtains solutions for the plasma density, electron temperature, and electric field for the given operating conditions. The physical phenomena and processes such as ambipolar diffusion, thermal diffusion, quasi-neutrality, ionization, inductive Joule heating, and excitations are considered in the model. A significant feature of the model is its capability of handling complex geometries that are often encountered in industrial reactors. Complex reactor geometries are modeled by a body-fitted-coordinate (BFC) formulation. A series of numerical experiments have been conducted using the model to study effects of various parameters such as chamber pressure, size of the wafer, position of the inductive coil, and the power input into the plasma. Different reactor geometries such as the GEC ICP reference cell and the belljar reactor have been simulated. The results of the parametric experiments are presented to show certain systematic trends in performance parameters such as uniformity and processing rates. The ICP model has been coupled to a computational fluid dynamics (CFD) code (capable of 3D simulations) that obtains the flow and pressure distribution inside the chamber. The ICP model will use pressure predictions (from the CFD model) to compute the local ionization rates. Chemical source/sink terms from the plasma dissociation model will be used by the CFD code to account for local reactant depletion effects.

  4. 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.

  5. 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.

  6. Optimization of Pressurized Oxy-Combustion with Flameless Reactor

    SciTech Connect

    Malavasi, Massimo; Landegger, Gregory

    2014-06-30

    Pressurized OxyECombustion is one of the most promising technologies for utility-scale power generation plants. Benefits include the ability to burn low rank coal and capture CO2. 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 100 kW 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 oxy-combustion, and a competitive cost of electricity plus greater CO2 avoidance costs then prior study’s presented. UPA’s operation of the bench-scale system provided evidence that key performance targets were achieved: flue gas sampled at the combustor outlet had non-detectable residual fly ashes, and low levels of SO3 and heavy-metal. These results correspond to prior pressurized oxy-combustion testing completed by IteaEEnel.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Cooling arrangement for the locking device in pressure gasification reactors

    SciTech Connect

    Jaschke, P.; Monch, E.; Weber, R.

    1981-06-02

    A cooling arrangement particularly for cooling a locking device arranged in a fuel feed opening of pressure gasification generators comprises a lock seat connected to the generator and a lock cone adapted for moving into and out of engagement with the lock seat and provided with a supply line to provide a cooling medium to the lock cone and a discharge line through which the used medium is drained off the lock cone. The lock cone is formed as a hollow body with a cooling chamber in its interior and is provided with guide means extending towards the lowermost area of the lock cone to direct the cooling medium into contact with the mostly overheated lowermost area of the lock cone positioned in a close relationship with the pressure gasification generator.

  15. 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)

  16. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    NASA Astrophysics Data System (ADS)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  17. Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Alipchenkov, V. M.; Belikov, V. V.; Davydov, A. V.; Emel'yanov, D. A.; Mosunova, N. A.

    2013-05-01

    Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

    PubMed

    Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

    2015-02-01

    Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. PMID:25522301

  3. Proceedings: 2002 Workshop on Pressurized Water Reactor Elevated Feedwater Iron Transport

    SciTech Connect

    2002-11-01

    Some pressurized water reactor (PWR) stations have experienced difficulty with maintaining feedwater (FW) iron concentrations below recommended concentration on a regular basis. A workshop held on September 17-18 in Dana Point, California, addressed the challenge of elevated feedwater iron transport in PWRs.

  4. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-20

    ... components exposed to treated borated water. DATES: Submit comments by May 21, 2012. Comments received after... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... availability was published in the Federal Register on June 22, 2010 (75 FR 35510). The NRC staff has...

  5. 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.

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

    PubMed

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

    2015-10-01

    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. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... document. You may access publicly-available information related to this action by the following methods... CONTACT section of this document. NRC's Agencywide Documents Access and Management System (ADAMS): You may... emergency core cooling systems (ECCSs) of pressurized water reactors (PWRs). This RG also describes...

  8. Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

    PubMed

    Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

    2015-02-01

    Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. A new scanning tunneling microscope reactor used for high-pressure and high-temperature catalysis studies.

    PubMed

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

    2008-08-01

    We present the design and performance of a homebuilt 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 ( approximately 19 cm(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(-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.

  16. 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.

  17. Annealing behaviour of reactor pressure-vessel steels studied by positron-annihilation spectroscopy, Mössbauer spectroscopy and transmission electron microscopy

    NASA Astrophysics Data System (ADS)

    Slugeǹ, V.; Segers, D.; de Bakker, P. M. A.; De Grave, E.; Magula, V.; Van Hoecke, T.; Van Waeyenberge, B.

    The annealing behaviour of commonly used reactor pressure-vessel steels was studied using positron-annihilation spectroscopy (PAS) (lifetime and Doppler broadening techniques), transmission Mössbauer spectroscopy (MS), integral low-energy electron MS (ILEEMS) and transmission electron microscopy (TEM). The results of these methods applied to different types of steels are discussed. Some significant differences in the experimental results are observed between the Russian (3 types) and Western Europe (5 types) steels and between individual specimens within these two groups. All specimens were annealed in vacuum and studied after this thermal treatment. It was confirmed that the heat affected zone (HAZ) is the most sensitive region for thermal and neutron irradiation-induced embrittlement in the reactor. Positron-annihilation lifetime measurements on HAZ specimens annealed at successively higher temperatures show the drastic increase in the vacancy-type defect formation between 525°C and 600°C. Therefore these specimens were selected for further detailed studies by TEM.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Advanced pressurized water reactor for improved resource utilization, part II - composite advanced PWR concept

    SciTech Connect

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

    1981-09-15

    This report evaluates the enhanced resource utilization in an advanced pressurized water reactor (PWR) concept using a composite of selected improvements identified in a companion study. The selected improvements were in the areas of reduced loss of neutrons to control poisons, reduced loss of neutrons in leakage from the core, and improved blanket/reflector concepts. These improvements were incorporated into a single composite advanced PWR. A preliminary assessment of resource requirements and costs and impact on safety are presented.

  4. 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).

  5. 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.

  6. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    SciTech Connect

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal.

  7. 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.

  8. 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

  9. 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.

  10. Radial pressure profiles in a cold‐flow gas‐solid vortex reactor

    PubMed Central

    Pantzali, Maria N.; Kovacevic, Jelena Z.; Marin, Guy B.; Shtern, Vladimir N.

    2015-01-01

    A unique normalized radial pressure profile characterizes the bed of a gas‐solid vortex reactor over a range of particle densities and sizes, solid capacities, and gas flow rates: 950–1240 kg/m3, 1–2 mm, 2 kg to maximum solids capacity, and 0.4–0.8 Nm3/s (corresponding to gas injection velocities of 55–110 m/s), respectively. The combined momentum conservation equations of both gas and solid phases predict this pressure profile when accounting for the corresponding measured particle velocities. The pressure profiles for a given type of particles and a given solids loading but for different gas injection velocities merge into a single curve when normalizing the pressures with the pressure value downstream of the bed. The normalized—with respect to the overall pressure drop—pressure profiles for different gas injection velocities in particle‐free flow merge in a unique profile. © 2015 The Authors AIChE Journal published by Wiley Periodicals, Inc. on behalf of American Institute of Chemical Engineers AIChE J, 61: 4114–4125, 2015 PMID:27667827

  11. Radial pressure profiles in a cold‐flow gas‐solid vortex reactor

    PubMed Central

    Pantzali, Maria N.; Kovacevic, Jelena Z.; Marin, Guy B.; Shtern, Vladimir N.

    2015-01-01

    A unique normalized radial pressure profile characterizes the bed of a gas‐solid vortex reactor over a range of particle densities and sizes, solid capacities, and gas flow rates: 950–1240 kg/m3, 1–2 mm, 2 kg to maximum solids capacity, and 0.4–0.8 Nm3/s (corresponding to gas injection velocities of 55–110 m/s), respectively. The combined momentum conservation equations of both gas and solid phases predict this pressure profile when accounting for the corresponding measured particle velocities. The pressure profiles for a given type of particles and a given solids loading but for different gas injection velocities merge into a single curve when normalizing the pressures with the pressure value downstream of the bed. The normalized—with respect to the overall pressure drop—pressure profiles for different gas injection velocities in particle‐free flow merge in a unique profile. © 2015 The Authors AIChE Journal published by Wiley Periodicals, Inc. on behalf of American Institute of Chemical Engineers AIChE J, 61: 4114–4125, 2015

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. First-principles investigation of boron incorporation into CRUD under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.

    2014-03-01

    CRUD (Chalk River Unidentified Deposit) is predominately a nickel-ferrite deposit on hot surfaces of nuclear fuel rods during reactor operation. The presence of CRUD modifies the core-coolant heat transfer and can induce localized corrosion on the cladding surface. Besides these unwanted effects boron, which is a neutron absorber, can accumulate within the CRUD, triggering shifts in the neutron flux and fluctuations in the reactor power level. Therefore, it is crucial to understand and predict the mechanisms by which B is trapped into the CRUD. As a first step, the incorporation of B defect into the crystal structure of NiFe2O4 has been investigated using the DFT framework. To obtain the formation energies of various interstitial and substitutional B-defects, theoretical results have been combined with experimental thermo-chemical data. Assuming solid-solid equilibrium conditions, the main factors that limit the incorporation of B are (i) the narrow stability domain of the host NiFe2O4 and (ii) the formation of ternary Fe-B-O and Ni-B-O compounds. The study also investigates the incorporation of B assuming solid-liquid equilibrium between NiFe2O4 and the surrounding aqueous solution under conditions of pressure, temperature, and pH characteristic to pressurized water reactors.

  18. 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.

  19. 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.

  20. The Gamma Ray Spectra Measurement in Reactor Pressure Vessel Simulator in WWER-1000 Engineering Benchmark

    NASA Astrophysics Data System (ADS)

    Ošmera, B.; Kyncl, J.; Mařík, M.; Cvachovec, F.; Smutný, V.; Králík, M.

    2003-06-01

    The space - energy distribution of the mixed neutron - photon radiation field has been measured over the Reactor Pressure Vessel (RPV) simulator thickness in the WWER-1000 engineering benchmark assembly in the LR-0 experimental reactor (in Nuclear Research Institute Řež plc) with a scintillation spectrometer. The spectra have been measured before the RPV in one quarter, one half, and three quarters of its thickness and behind the RPV in the energy range 0.5 - 10 MeV. The evaluated integral fluxes above 1 MeV and their ratio are compared with the MCNP and DORT calculation, the measured spectra are presented graphically. The measurements are being performed in the frame of the project REDOS [1], 5th Framwork Programme of the European Community 1998 - 2002.

  1. The French neutronic program addressing the requirements of future pressurized water reactors

    SciTech Connect

    Bergeron, J.; Darrouzet, M.; Gomit, J.M.; Lenain, R.; Nigon, J.L.; Martin-Deidier, L.

    1988-02-01

    Undermoderated plutonium lattices in pressurized water reactors (PWRs) offer the prospect of very significant uranium savings as well as less degradation of the plutonium isotopic quality. There are, indeed, some uncertainties associated with such designs and, in 1984, an extensive program was launched to reduce these uncertainties. For the neutronic reactor physics research, the program included: (a) adaptation of codes - in particular, the cell code APOLLO and its neutronic data library; (b) an extensive experimental program (the ERASME, ICARE, and MORGANE experiments) to reduce the uncertainties in the different neutronic parameters; and (c) studies of important and specific future PWR problems. This program allows us to qualify the neutronic codes for the undermoderated lattices in a very large range of moderator ratios and to validate future PWR designs.

  2. Automated Microdosing System for Integration With a Miniaturized High-pressure Reactor System

    PubMed Central

    Hawali, Ihsan; Thurow, Kerstin

    2005-01-01

    We present a new automated dosing system developed by the Institute for Automation of the University of Rostock, Germany. The new system is designed for the dosing of chemical liquids in the range of 50 μL–2.5 mL. It is integrated into a miniaturized reactor system to be used in the field of combinatorial synthesis. The reactor system can be pressurized up to 150 bar and tempered up to 200^C. A wide range of liquids with different physical properties can be handled with the new dosing system. A detailed description of the new dosing system in terms of function and operation as well as the relevant features and potential benefits is provided. PMID:18924738

  3. Automated Microdosing System for Integration With a Miniaturized High-pressure Reactor System.

    PubMed

    Stoll, Norbert; Hawali, Ihsan; Thurow, Kerstin

    2005-01-01

    We present a new automated dosing system developed by the Institute for Automation of the University of Rostock, Germany. The new system is designed for the dosing of chemical liquids in the range of 50 muL-2.5 mL. It is integrated into a miniaturized reactor system to be used in the field of combinatorial synthesis. The reactor system can be pressurized up to 150 bar and tempered up to 200(;)C. A wide range of liquids with different physical properties can be handled with the new dosing system. A detailed description of the new dosing system in terms of function and operation as well as the relevant features and potential benefits is provided.

  4. 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.

  5. Thorium plutonium (TREX) fuel for weapons-grade plutonium disposition in pressurized water reactors

    SciTech Connect

    Comfort, C.; Ferguson, C.; Klima, S.; Lilly, D.E.; Rahnema, F.

    1996-12-31

    The goal of this study was to create a pressurized water reactor (PWR) reactor assembly (17 x 17) that would burn weapons-grade plutonium (WGP). Current designs of mixed-oxide (MOX) fuels combine WGP with uranium as the fuel. MOX fuel assemblies will destroy plutonium, but only 40 to 50% of the plutonium present in the fuel. This percentage is limited by the presence of {sup 238}U in the core, which becomes {sup 239}Pu by absorption and decay. The production of plutonium counteracts the disposition of WGP in current MOX fuel designs. This problem can be overcome by replacing the uranium in a MOX design with thorium. This loss of uranium (primarily {sup 238}U) halts the production of {sup 239}Pu in the thorium plutonium (TREX) fuel. The absence of {sup 239}Pu production allows the TREX design to burn up to 85 wt% of the {sup 239}Pu, originally loaded in the fuel.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

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

    DOE PAGES

    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

  11. Atomic layer deposition of Al(2)O(3) and ZnO at atmospheric pressure in a flow tube reactor.

    PubMed

    Jur, Jesse S; Parsons, Gregory N

    2011-02-01

    Improving nanoscale thin film deposition techniques such as atomic layer deposition (ALD) to permit operation at ambient pressure is important for high-throughput roll-to-roll processing of emerging flexible substrates, including polymer sheets and textiles. We present and investigate a novel reactor design for inorganic materials growth by ALD at atmospheric pressure. The reactor uses a custom "pressure boost" approach for delivery of low vapor pressure ALD precursors that controls precursor dose independent of reactor pressure. Analysis of continuum gas flow in the reactor shows key relations among reactor pressure, inert gas flow rate, and species diffusion that define conditions needed to efficiently remove product and adsorbed reactive species from the substrate surface during the inert gas purge cycle. Experimental results, including in situ quartz crystal microbalance (QCM) characterization and film thickness measurements for deposition of ZnO and Al(2)O(3) are presented and analyzed as a function of pressure and gas flow rates at 100 °C. At atmospheric pressure and high gas flow, ZnO deposition can proceed at the same mass uptake and growth rate as observed during more typical low pressure ALD. However, under the same high pressure and flow conditions the mass uptake and growth rate for Al(2)O(3) is a factor of ∼1.5-2 larger than at low pressure. Under these conditions, Al(2)O(3) growth at atmospheric pressure in a "flow-through" geometry on complex high surface area textile materials is sufficiently uniform to yield functional uniform coatings.

  12. 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).

  13. 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

  14. 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.

  15. A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

    SciTech Connect

    Witte, M.C.; Chou, C.K.

    1989-03-24

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs.

  16. 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.

  17. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    SciTech Connect

    Smith, Curtis; Mandelli, Diego; 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 impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Irradiation Induced Defect Characterization in Reactor Pressure Vessel Steel by Small Angle Neutron Scattering

    NASA Astrophysics Data System (ADS)

    Han, Yougn-Soo; Shin, Eun-Joo; Lee, Chang-Hee; Park, Duck-Gun

    The degradation of the mechanical properties of the RPV (Reactor Pressure Vessel) steel during an irradiation in a nuclear power plant is closely related to the irradiation induced defects. The size of these defects is known to be a few nanometer, and the small angle neutron scattering technique is regarded as the best non destructive technique to characterize the nano sized inhomogeneities in bulk samples. The investigated the RPV steel has been used in YeongKwang nuclear power plant at Korea and the Cu content of the RPV steel is 0.06 wt%. The RPV steel was irradiated in the HANARO reactor at KAERI. The small angle neutron scattering experiments were performed by the SANS instrument in the HANARO reactor. The nano sized irradiation induced defects were quantitatively analyzed by SANS and the type of the irradiation induced defects was discussed in detail. The relation between irradiation induced defects and the yield strength was investigated. The characteristics of irradiation induced defects in low Cu containing RPV steel were discussed.

  3. 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

  4. 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

  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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    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. 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.

  8. A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy

    SciTech Connect

    Rozon, Daniel; Shen Wei

    2001-05-15

    For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

  9. 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.

  10. Rapid pressure swing absorption cleanup of post-shift reactor synthesis gas

    SciTech Connect

    Sirkar, K.K.; Majumdar, S.; Bhaumik, S.

    1991-10-31

    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. The specific objectives are (1) to fiber membranes. The specific objectives are (1) to develop a theoretical model which will provide guidelines for selecting an efficient RAPSAB process cycle which includes desorption; (2) to demonstrate the concept experimentally with a simple gas mixture (e.g., Co{sub 2}-N{sub 2}) and a simple absorbent liquid such as water, and (3) to extend the concept to reactive absorbent liquids for the separation of CO, Co{sub 2} from the post-shift reactor synthesis gas. A simplified theoretical description of the novel rapid pressure swing absorption process has been developed. The absorption part of the pressure swing absorption cycle has been predicted for CO{sub 2}-N{sub 2}-water system. Numerical simulation of the model is being carried out for different operating conditions for selecting an optimum pressure swing cycle.

  11. 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.

  12. 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)

  13. Russian roulette

    NASA Astrophysics Data System (ADS)

    Showstack, Randy

    Following a recent collision, fire, series of computer and power failures, and other mishaps on the Russian space station, Mir, the U.S. Congress held a hearing on September 18 to question the safety of American astronauts staying aboard the aging spacecraft.“There has been sufficient evidence put before this hearing to raise doubts about the safety of continued American long-term presence on the Mir,” said House Science Committee Chairman Rep. James Sensenbrenner (R-Wisc.) at the hearing.

  14. Oxidation of automotive primary reference fuels in a high pressure flow reactor

    SciTech Connect

    Curran, H.J.; Pitz, W.J.; Westbrook, C.K.; Callahan, C.V.; Dryer, F.L.

    1998-01-01

    Automotive engine knock limits the maximum operating compression ratio and ultimate thermodynamic efficiency of spark-ignition (SI) engines. In compression-ignition (CI) or diesel cycle engines the premixed urn phase, which occurs shortly after injection, determines the time it takes for autoignition to occur. In order to improve engine efficiency and to recommend more efficient, cleaner-burning alternative fuels, we must understand the chemical kinetic processes which lead to autoignition in both SI and CI engines. These engines burn large molecular-weight blended fuels, a class to which the primary reference fuels (PRF), n-heptane and isooctane belong. In this study, experiments were performed under engine-like conditions in a high pressure flow reactor using both the pure PRF fuels and their mixtures in the temperature range 550-880 K and at 12.5 atm pressure. These experiments not only provide information on the reactivity of each fuel but also identify the major intermediate products formed during the oxidation process. A detailed chemical kinetic mechanism is used to simulate these experiments and comparisons of experimentally measures and model predicted profiles for O{sub 2}, CO, CO{sub 2}, H{sub 2}O and temperature rise are presented. Intermediates identified in the flow reactor are compared with those present in the computations, and the kinetic pathways leading to their formation are discussed. In addition, autoignition delay times measured in a shock tube over the temperature range 690- 1220 K and at 40 atm pressure were simulated. Good agreement between experiment and simulation was obtained for both the pure fuels and their mixtures. Finally, quantitative values of major intermediates measured in the exhaust gas of a cooperative fuels research engine operating under motored engine conditions are presented together with those predicted by the detailed method.

  15. Feasibility for development of a nuclear reactor pressure vessel flaw distribution: Sensitivity analyses and NDE (nondestructive evaluation) capability

    SciTech Connect

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

    1990-01-01

    Pressurized water reactor pressure vessels operate under US Nuclear Regulatory Commission (NRC) rules and regulatory guides that are intended to maintain a low probability of vessel failure. The NRC has also addressed neutron embrittlement of pressurized water reactor pressure vessels by imposing regulations on plant operation. Plants failing to meet the operating criteria specified by these rules and regulations are required, among other things, to analytically demonstrate fitness for service in order to continue safe operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. A fracture mechanics sensitivity study was performed to quantify the effect of the assumed flaw distribution on the predicted vessel performance under a specified pressurized thermal shock transient and to determine the critical crack size. Results of the analysis indicate that vessel performance in terms of the estimated probability of failure is very sensitive to the assumed flaw distribution. 20 refs., 3 figs., 2 tabs.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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

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

    DOE PAGES

    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

  3. 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.

  4. BWR reactor pressure vessel license renewal industry report; revision 1. Final report

    SciTech Connect

    Braden, D.; Stancavage, P.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) Reactor Pressure Vessels (RPVs). The report includes the following RPV components: attachment welds, closure studs, nozzles and safe ends, penetrations, vessel shell and flanges, top and bottom heads and vessel support skirt. The scope is limited to domestic BWRs designated by GE Nuclear Energy as BWR/2 through BWR/6.

  5. 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.

  6. JRQ and JPA irradiated and annealed reactor pressure vessel steels studied by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Gokhman, Oleksandr; Pecko, Stanislav; Sojak, Stanislav; Bergner, Frank

    2016-03-01

    The paper is focused on a comprehensive study of JRQ and JPA reactor pressure vessel steels from the positron annihilation lifetime spectroscopy (PALS) point of view. Based on our more than 20 years' experience with characterization of irradiated reactor steels, we confirmed that defects after irradiation start to grow and/or merge into bigger clusters. Experimental results shown that JPA steel is more sensitive to the creation of irradiation-induced defects than JRQ steel. It is most probably due to high copper content (0.29 wt.% in JPA) and copper precipitation has a major impact on neutron-induced defect creation at the beginning of the irradiation. Based on current PALS results, no large vacancy clusters were formed during irradiation, which could cause dangerous embrittlement concerning operation safety of nuclear power plant. The combined PALS, small angle neutron scattering and atomic probe tomography studies support the model for JRQ and JPA steels describing the structure of irradiation-induced clusters as agglomerations of vacancy clusters (consisting of 2-6 vacancies each) and are separated from each other by a distribution of atoms.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Development of an improved-contact liquid-level probe for pressurized reactor vessels

    SciTech Connect

    Kelsey, P.V. Jr.

    1982-09-01

    Electrical-conductivity-based probes for liquid level sensing show promise for pressurized water reactor environments, but have exhibited frequent bond failures at the ceramic/metal interfaces. A program to characterize and improve the interface behavior has been completed successfully, and has provided data for optimizing fabrication parameters, as well as general information on glass-to-metal bonding in a superalloy/silicate-glass system. The materials studied were Inconel X-750 and a barium silicate glass containing minor amounts of TiO/sub 2/, CeO/sub 2/, As/sub 2/O/sub 3/, Bi/sub 2/O/sub 3/, and Al/sub 2/O/sub 3/.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 14C release from a Soviet-designed pressurized water reactor nuclear power plant.

    PubMed

    Uchrin, G; Csaba, E; Hertelendi, E; Ormai, P; Barnabas, I

    1992-12-01

    The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

  18. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    NASA Astrophysics Data System (ADS)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  19. 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.

  20. High Throughput Atomic Layer Deposition Processes: High Pressure Operations, New Reactor Designs, and Novel Metal Processing

    NASA Astrophysics Data System (ADS)

    Mousa, MoatazBellah Mahmoud

    Atomic Layer Deposition (ALD) is a vapor phase nano-coating process that deposits very uniform and conformal thin film materials with sub-angstrom level thickness control on various substrates. These unique properties made ALD a platform technology for numerous products and applications. However, most of these applications are limited to the lab scale due to the low process throughput relative to the other deposition techniques, which hinders its industrial adoption. In addition to the low throughput, the process development for certain applications usually faces other obstacles, such as: a required new processing mode (e.g., batch vs continuous) or process conditions (e.g., low temperature), absence of an appropriate reactor design for a specific substrate and sometimes the lack of a suitable chemistry. This dissertation studies different aspects of ALD process development for prospect applications in the semiconductor, textiles, and battery industries, as well as novel organic-inorganic hybrid materials. The investigation of a high pressure, low temperature ALD process for metal oxides deposition using multiple process chemistry revealed the vital importance of the gas velocity over the substrate to achieve fast depositions at these challenging processing conditions. Also in this work, two unique high throughput ALD reactor designs are reported. The first is a continuous roll-to-roll ALD reactor for ultra-fast coatings on porous, flexible substrates with very high surface area. While the second reactor is an ALD delivery head that allows for in loco ALD coatings that can be executed under ambient conditions (even outdoors) on large surfaces while still maintaining very high deposition rates. As a proof of concept, part of a parked automobile window was coated using the ALD delivery head. Another process development shown herein is the improvement achieved in the selective synthesis of organic-inorganic materials using an ALD based process called sequential vapor

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

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

  2. 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)…

  3. 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.

  4. 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.

  5. 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.

  6. A perspective on thermal annealing of reactor pressure vessel materials from the viewpoint of experimental results

    SciTech Connect

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

    1996-04-01

    It is believed that in the next decade or so, several nuclear reactor pressure vessels (RPVs) may exceed the reference temperature limits set by the pressurized thermal shock screening criteria. One of the options to mitigate the effects of irradiation on RPVs 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 to study the annealing response, or ``recovery`` of several irradiated RPV steels. The fracture toughness is one of the important properties used in the evaluation of the integrity of RPVs. Optimally, the fracture toughness is measured directly by fracture toughness specimens, such as compact tension or precracked Charpy specimens, but is often inferred from the results of Charpy V-notch impact specimens. The experimental results are compared to the predictions of models for embrittlement recovery which have been developed by Eason et al. Some of the issues in annealing that still need to be resolved are discussed.

  7. 77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-03

    ... Information The draft SFST-ISG-8, Revision 3, was published in the Federal Register on May 2, 2012 (77 FR... safety analyses of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and...) optional credit for fission product and minor actinide neutron absorbing isotopes in the SNF...

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

    NASA Astrophysics Data System (ADS)

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

    2011-05-01

    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

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Three somatic genetic biomarkers and covariates in radiation-exposed Russian cleanup workers of the chernobyl nuclear reactor 6-13 years after exposure.

    PubMed

    Jones, Irene M; Galick, Heather; Kato, Paula; Langlois, Richard G; Mendelsohn, Mortimer L; Murphy, Gloria A; Pleshanov, Pavel; Ramsey, Marilyn J; Thomas, Cynthia B; Tucker, James D; Tureva, Ludmila; Vorobtsova, Irina; Nelson, David O

    2002-10-01

    Three somatic mutation assays were evaluated in men exposed to low-dose, whole-body, ionizing radiation. Blood samples were obtained between 1992 and 1999 from 625 Russian Chernobyl cleanup workers and 182 Russian controls. The assays were chromosome translocations in lymphocytes detected by FISH, hypoxanthine phosphoribosyltransferase (HPRT) mutant frequency in lymphocytes by cloning, and flow cytometic assay for glycophorin A (GPA) variant frequency of both deletion (N/Ø) and recombination (N/N) events detected in erythrocytes. Over 30 exposure and lifestyle covariates were available from questionnaires. Among the covariates evaluated, some increased (e.g. age, smoking) and others decreased (e.g. date of sample) biomarker responses at a magnitude comparable to Chernobyl exposure. When adjusted for covariates, exposure at Chernobyl was a statistically significant factor for translocation frequency (increase of 30%, 95% CI of 10%-53%, P = 0.002) and HPRT mutant frequency (increase of 41%, 95% CI of 19%-66%, P < 0.001), but not for either GPA assay. The estimated average dose for the cleanup workers based on the average increase in translocations was 9.5 cGy. Translocation analysis is the preferred biomarker for low-dose radiation dosimetry given its sensitivity, relatively few covariates, and dose-response data. Based on this estimated dose, the risk of exposure-related cancer is expected to be low.

  15. 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

    ... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; Request for public... Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-4029..., Division of License Renewal, Office of Nuclear Reactor Regulation. BILLING CODE 7590-01-P...

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Neutron induced damage in reactor pressure vessel steel: An X-ray absorption fine structure study

    NASA Astrophysics Data System (ADS)

    Kuri, G.; Cammelli, S.; Degueldre, C.; Bertsch, J.; Gavillet, D.

    2009-03-01

    The radiation damage produced in reactor pressure vessel (RPV) steels during neutron irradiation is a long-standing problem of considerable practical interest. In this study, an extended X-ray absorption fine structure (EXAFS) spectroscopy has been applied at Cu, Ni and Mn K-edges to systematically investigate neutron induced radiation damage to the metal-site bcc structure of RPV steels, irradiated with neutrons in the fluence range from 0.85 to 5.0 × 1019 cm-2. An overall similarity of Cu, Ni and Mn atomic environment in the iron matrix is observed. The radial distribution functions (RDFs), derived from EXAFS data have been found to evolve continuously as a function of neutron fluence describing the atomic-scale structural modifications in RPVs by neutron irradiations. From the pristine data, long range order beyond the first- and second-shell is apparent in the RDF spectra. In the irradiated specimens, all near-neighbour peaks are greatly reduced in magnitude, typical of damaged material. Prolonged annealing leads annihilation of point defects to give rise to an increase in the coordination numbers of near-neighbour atomic shells approaching values close to that of non-irradiated material, but does not suppress the formation of nano-sized Cu and/or Ni-rich-precipitates. Total amount of radiation damage under a given irradiation condition has been determined. The average structural parameters estimated from the EXAFS data are presented and discussed.

  3. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    SciTech Connect

    Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott; Hsu, C.

    1999-08-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  4. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

    SciTech Connect

    A. G. Ware; C. Hsu; C. L. Atwood; M. B. Sattison; R. S. Hartley; V. N. Shah

    1999-02-01

    This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

  5. 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.

  6. 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.

  7. 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.

  8. Aging mechanisms in the Westinghouse PWR (Pressurized Water Reactor) Control Rod Drive system

    SciTech Connect

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs.

  9. 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

  10. 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.

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

    DOE PAGES

    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

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

    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 fuel 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. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan

    SciTech Connect

    Yoshiaki Oka; Katsumi Yamada

    2004-07-01

    This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

  3. 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

    , ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)

  4. 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.

  5. 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.

  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. KWU's high conversion reactor concept - An economical evolution of modern pressurized water reactor technology toward improved uranium ore utilization

    SciTech Connect

    Markl, H.; Goetzmann, C.A.; Moldaschl, H.

    1988-01-01

    The Kraftwerk Union AG high conversion reactor represents a quasi-standard PWR with fuel assemblies of more or less uniformly enriched fuel rods, arranged in a tight hexagonal array with a pitch-to-diameter ratio p/d approx. = 1.12. High fuel enrichment as well as a high conversion ratio of --0.9 will provide the potential for high burnup values up to 70 000 MWd/tonne and a low fissile material consumption. The overall objective of the actual RandD program is to have the technical feasibility, including that for licensibility, established by the early 1990s as a prerequisite for deciding whether to enter a demonstration plant program.

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. Global threat reduction initiative Russian nuclear material removal progress

    SciTech Connect

    Cummins, Kelly

    2008-07-15

    In December 1999 representatives from the United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) started discussing a program to return to Russia Soviet- or Russian-supplied highly enriched uranium (HEU) fuel stored at the Russian-designed research reactors outside Russia. Trilateral discussions among the United States, Russian Federation, and the International Atomic Energy Agency (IAEA) have identified more than 20 research reactors in 17 countries that have Soviet- or Russian-supplied HEU fuel. The Global Threat Reduction Initiative's Russian Research Reactor Fuel Return Program is an important aspect of the U.S. Government's commitment to cooperate with the other nations to prevent the proliferation of nuclear weapons and weapons-usable proliferation-attractive nuclear materials. To date, 496 kilograms of Russian-origin HEU have been shipped to Russia from Serbia, Latvia, Libya, Uzbekistan, Romania, Bulgaria, Poland, Germany, and the Czech Republic. The pilot spent fuel shipment from Uzbekistan to Russia was completed in April 2006. (author)

  17. 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

  18. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

    DOE PAGES

    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

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. High glucose selectivity in pressurized water hydrolysis of cellulose using ultra-fast reactors.

    PubMed

    Cantero, Danilo A; Dolores Bermejo, M; José Cocero, M

    2013-05-01

    A new reactor was developed for the selective hydrolysis of cellulose. In this study, the glucose selectivity obtained from cellulose was improved by using ultra-fast reactions in which a selective medium was combined with an effective residence time control. A selective production of glucose, fructose and cellobiose (50%) or total mono-oligo saccharides (>96%) was obtained from the cellulose in a reaction time of 0.03 s. Total cellulose conversion was achieved with a 5-hydroxymethylfural concentration lower than 5 ppm in a novel micro-reactor. Reducing the residence time from minutes to milliseconds opens the possibility of moving from the conventional m(3) to cm(3) reactor volumes.

  5. 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.

  6. 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.

  7. 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.

  8. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased. Comparing cluster IDs 2 and 6, embrittlement of high-Cu-bearing materials (<0.07 wt%) was larger than that of low-Cu-bearing (0.07 < wt.%) materials. This is attributed to irradiation-induced Cu-enriched clusters, as well as those that are irradiation-enhanced [4]. A similar feature is recognized for cluster IDs 5 and 8 in materials with a higher Ni content. A flux effect with a higher flux range was demonstrated for cluster ID 3 comprising MTR irradiation in a high flux region (≤1 × 1013 n/cm2/s) [44]. For cluster ID 10, classification is rendered based upon flux effect, where embrittlement is accelerated in high Cu-bearing materials irradiated at lower flux levels (less than 5 × 109 n/cm2·s). This is possibly due to increased thermal equilibrium vacancies [44,45]. Per all the above considerations, it was hence ascertained that data belonging to identical cluster ID

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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…

  15. 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…

  16. 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…

  17. Prediction of Severe Accident Counter Current Natural Circulation Flows in the Hot Leg of a Pressurized Water Reactor

    SciTech Connect

    Boyd, Christopher F.

    2006-07-01

    During certain phases of a severe accident in a pressurized water reactor (PWR), the core becomes uncovered and steam carries heat to the steam generators through natural circulation. For PWR's with U-tube steam generators and loop seals filled with water, a counter current flow pattern is established in the hot leg. This flow pattern has been experimentally observed and has been predicted using computational fluid dynamics (CFD). Predictions of severe accident behavior are routinely carried out using severe accident system analysis codes such as SCDAP/RELAP5 or MELCOR. These codes, however, were not developed for predicting the three-dimensional natural circulation flow patterns during this phase of a severe accident. CFD, along with a set of experiments at 1/7. scale, have been historically used to establish the flow rates and mixing for the system analysis tools. One important aspect of these predictions is the counter current flow rate in the nearly 30 inch diameter hot leg between the reactor vessel and steam generator. This flow rate is strongly related to the amount of energy that can be transported away from the reactor core. This energy transfer plays a significant role in the prediction of core failures as well as potential failures in other reactor coolant system piping. CFD is used to determine the counter current flow rate during a severe accident. Specific sensitivities are completed for parameters such as surge line flow rates, hydrogen content, as well as vessel and steam generator temperatures. The predictions are carried out for the reactor vessel upper plenum, hot leg, a portion of the surge line, and a steam generator blocked off at the outlet plenum. All predictions utilize the FLUENT V6 CFD code. The volumetric flow in the hot leg is assumed to be proportional to the square root of the product of normalized density difference, gravity, and hydraulic diameter to the 5. power. CFD is used to determine the proportionality constant in the range

  18. The ReactorSTM: atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions.

    PubMed

    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.

  19. 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.

  20. 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.

  1. 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.

  2. The ReactorSTM: atomically resolved scanning tunneling microscopy under high-pressure, high-temperature catalytic reaction conditions.

    PubMed

    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. PMID:25173272

  3. 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.

  4. 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.

  5. 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.

  6. Analysis and selection of high pressure heaters design for a new generation of NPP with BN-1200 reactor plant

    NASA Astrophysics Data System (ADS)

    Yurchenko, A. Yu.; Sukhorukov, Yu. G.; Trifonov, N. N.; Grigor'eva, E. B.; Esin, S. B.; Svyatkin, F. A.; Nikolaenkova, E. K.; Prikhod'ko, P. Yu.; Nazarov, V. V.

    2016-09-01

    In the development of advanced high-power steam-turbine plants (STP), special attention is placed on the design of reliable and economical high-pressure heater (HPH) capable to maintain the specified thermal hydraulic performance during the entire service life. Comparative analysis of the known designs of HPH, such as the spiral-collector HPH, the collector-coiled HPH, the collector-platen HPH, modular HPH, and the chamber HPH, was carried out. The advantages and disadvantages of each design were pointed. For better comparison, the heaters are separated into two groups—horizontal and vertical ones. The weight and dimension characteristics, the materials and features of the basic elements, and operating features of variety HPH are presented. At operating the spiral-collector HPH used in the majority of regenerative schemes of high-pressure STP of thermal and nuclear power plants, the disadvantages reducing the economy and reliability of their operation were revealed. The recommendations directed to the reliability growth of HPH, the decrease of subcooling the feed water, the increase of compactness are stated. Some of these were developed by the specialists of OAO NPO TsKTI and are successfully implemented on the thermal power plants and nuclear power plants. Technical solutions to reduce the cost of regeneration system and the weight of chamber HPH, reduce the thickness of the tube plate of HPH, and reliability assurance of the cooler of steam and condensate built in the HPH casing under all operating conditions were proposed. Three types of feed water chambers for vertical and horizontal chamber HPH are considered in detail, the constructive solutions that have been implemented in HPH of the regeneration system of turbines of 1000 and 1200 MW capacity with water-moderated water-cooled power reactor (WMWCPR) are described. The optimal design of HPH for the regeneration system of high-pressure turbine plant with BN-1200 reactor was selected.

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. Small-angle neutron scattering investigation of as-irradiated, annealed and reirradiated reactor pressure vessel weld material of decommissioned reactor

    NASA Astrophysics Data System (ADS)

    Ulbricht, A.; Altstadt, E.; Bergner, F.; Viehrig, H.-W.; Keiderling, U.

    2011-09-01

    Small-angle neutron scattering (SANS) was applied to characterize the microstructure of weld material taken from the reactor pressure vessel (RPV) of the decommissioned VVER440 (230)-type nuclear power plant (NPP) Greifswald, Units 1, 2 and 4. The welding seam of highest neutron exposure of Unit 1 was subject to a large-scale annealing treatment in 1988 after about 11.5 effective years of operation. The same type of annealing was applied to Unit 2 in 1990 after about 11 effective years of operation. After final decommissioning of NPP Greifswald in 1990, RPV material was left in the reirradiated condition (Unit 1), in the as-annealed condition (Unit 2) and in the as-irradiated condition (Unit 4). Trepans of material from the highly irradiated RPV welds of these Units have recently become available for examination. The results of the SANS investigation are reported and compared with published results obtained for as-irradiated, post-irradiation annealed and reirradiated surveillance material of the same type. A general agreement was found indicating in particular the formation of irradiation-induced Cu-enriched clusters and efficient recovery as a result of the large-scale annealing treatments. The only essential difference was observed for the ratio of magnetic and nuclear scattering indicating differences of the cluster composition for the RPV wall and surveillance material.

  12. 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.

  13. Dependence of the growth rate of an AlN layer on nitrogen pressure in a reactor for sublimation growth of AlN crystals

    SciTech Connect

    Wolfson, A. A. Mokhov, E. N.

    2010-10-15

    The dependence of the layer growth rate on nitrogen pressure in a reactor has been examined in order to analyze the conditions of growth of AlN thick layers and bulk crystals by the sublimation sandwich method. It is shown that the layer growth rate steadily increases as the pressure in the reactor is lowered within the range 1-0.02 bar. This suggests that a key role in the layer growth kinetics is played by the source-to-substrate transfer of the components (Al, N), rather than by their adsorption (desorption) on the substrate surface.

  14. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    SciTech Connect

    Licht, J.; Bergeron, A.; Dionne, B.; Sikik, E.; Van den Branden, G.; Koonen, E.

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration (Figure 1). The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration (Figure 2), based on current reactor use, has been defined for the fuel conversion analyses [1]. The code RELAP5/Mod 3.3 [2] was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  15. 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.

  16. RBMK pressure tube rupture assessment

    SciTech Connect

    Schmitt, B.E.; Tsiklauri, G.V.

    1994-08-01

    The Russian RBMK reactor core design consists of multiple parallel pressure tube channels that contain Zr clad, UO{sub 2} fuel pin bundles. These parallel channels are contained within graphite moderator blocks which are, in turn, contained within a sealed core cavity. Current safety evaluation efforts of the RBMK reactors have been concentrating in the area of tube ruptures within the core cavity and, in particular, multiple tube ruptures that could threaten the reactor core integrity. Tube rupture events result in a pressurization of the reactor core cavity. The original design overpressure for the cavity region was based on a single tube rupture, resulting in considerable margin to the top plate lift pressure. The top plate lift pressure is 3.1 bar, and a single tube rupture would result in approximately 1.4 bar. RBMK plant specific cavity pressure relief designs provide for between three and in simultaneous tube ruptures before exceeding the top plate lift pressure. Thus, current safety evaluations have begun to examine the potential for multiple tube ruptures that could exceed the current cavity pressure relief designs. One such scenario being examined is a partial rupture in a group distribution header that results in stagnated (low) flow to up to 40 pressure tubes. The subsequent fuel heatup in these reduced flow tubes could result in multiple tube ruptures beyond the design relief capacity of the core cavity. This paper examines several key issues in evaluating this transient, including: (1) the effects of low flow, (2) the effects of axial peaking, and (3) the effects of radial peaking, all relative to the time to tube rupture. These issues each play a significant role in attempting to evaluate the likelihood and severity of multiple tube ruptures for a partial group distribution header break.

  17. Quantum Russian roulette

    NASA Astrophysics Data System (ADS)

    Schmidt, Alexandre G. M.; da Silva, Ladário

    2013-01-01

    We quantize the gamble known as Russian roulette and we study it in two versions for two- and three-persons when: (i) players use a fully loaded quantum gun; (ii) the quantum gun has only one quantum bullet.

  18. Russian Education: Historical Perspectives.

    ERIC Educational Resources Information Center

    Reynolds, John C.

    1981-01-01

    Discusses the evaluation of Russian educational concepts through six historical periods; stresses the significance of social, philosophical, political, and psychological factors that have affected Soviet educational policies used to instill Communist values. (JD)

  19. 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.

  20. 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...

  1. 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.

  2. 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.

  3. Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

    NASA Astrophysics Data System (ADS)

    Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

    2013-04-01

    In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

  4. 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.

  5. Effect of heat treatment on stress corrosion of Alloy 718 in pressurized-water-reactor primary water

    SciTech Connect

    Miglin, M.T.; Monter, J.V.; Wade, C.S.; Nelson, J.L.

    1992-12-31

    Stress corrosion cracking (SCC) tests were conducted in 360{degrees}C pressurized-water-reactor (PWR) primary water using alloy 718 in various heat treatment conditions. Alloy X-750 in the HTH condition and an experimental heat of an alloy 718 variation, Hicoroy, were also tested for comparison. Fatigue-precracked, 12.5-mm-thick compact fracture specimens were subjected to a constant extension rate of 1.3 x 10{sup {minus}9} m/s. Crack growth rate was measured during testing using a reversing DC potential drop technique. Results in the form of SCC crack growth rate versus applied stress intensity demonstrate that the SCC resistance of alloy 718 in the PWR primary-side environment can be improved by changes in heat treatment.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. The ReactorAFM: non-contact atomic force microscope operating under high-pressure and high-temperature catalytic conditions.

    PubMed

    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.

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. Analysis of operation of filters for post-accident decontamination of pressurized rooms of a nuclear power plants with a type VVER-440 reactor

    NASA Astrophysics Data System (ADS)

    Zaichik, L. I.; Zeigarnik, Yu. A.; Rotinov, A. G.; Sidorov, A. S.; Silina, N. N.; Chalyi, R. F.

    2007-05-01

    Operation of filters of postaccident decontamination of pressurized rooms of a nuclear power plant with a type-VVER-440 reactor is analyzed. The distribution of radioactive nuclides over filter stages, the time variation of the thermal state of filter, and the characteristic features of the processes of sorption in the section of fine cleaning are considered.

  16. 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.

  17. 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.

  18. Comparison of SERPENT and CASMO-5M for pressurized water reactors models

    SciTech Connect

    Hursin, M.; Vasiliev, A.; Ferroukhi, H.; Pautz, A.

    2013-07-01

    The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

  19. 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)

  20. 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.

  1. 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.

  2. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    SciTech Connect

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-07-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab.

  3. 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.

  4. 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

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. A Analytical and Experimental Investigation of Natural Circulation Transients in a Model Pressurized Water Reactor

    NASA Astrophysics Data System (ADS)

    Massoud, Mahmoud

    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 head. 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 behaviour to variation of inputs such as secondary distribution and power transients.

  10. 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…

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. Soviet Standardization of Russian.

    ERIC Educational Resources Information Center

    Benson, Morton

    1961-01-01

    After a brief history of influences on the standardization of Russian since the Revolution, the descriptive and normalizing role of specific reference works of grammar, phonetics, vocabulary, and morphology is discussed. Concluding remarks point out a problem caused by a lack of coordination of standardization efforts and the mounting demands for…

  19. Russian Resource Materials Guide.

    ERIC Educational Resources Information Center

    Seymour, Viveca

    This guide provides the teacher of Russian with helpful background material and activities on the geography of the Soviet Union and the history of Russia as well as its customs, traditions, literary selections, songs, foods, and festivals. In addition to these topics, the introductory chapter outlines a philosophy for teaching and learning Russian…

  20. Three Early Russian Bibliographers

    ERIC Educational Resources Information Center

    Cholden, Marianna Tax

    1974-01-01

    Introduces three early Russian contributors to bibliography: Novikov, Sopikov, and Anastasevich, who worked at the end of the 18th and beginning of the 19th centuries. Their lives and main bibliographic works are viewed in the contexts of Russia in their time and Western European bibliography. (JB)

  1. English Loanwords in Russian.

    ERIC Educational Resources Information Center

    Benson, Morton

    1959-01-01

    This introductory survey describes the English contribution to the vocabulary of modern Russian. The author presents an analysis of English loanwords based on the etymologies of Vasmer and Lexin, by subject classification (amusements, clothing, economics, food, nautical terminology, and technology). Separate commentary on sporting terms, where…

  2. 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

  3. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    DOE PAGES

    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

  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. 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.

  12. 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.

  13. 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.

  14. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    NASA Astrophysics Data System (ADS)

    Oh, Yong-Jun; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T0 values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 μm in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness ( KJC) was inversely proportional to the square root of the triggering inclusion diameter ( di) at respective temperatures. From this relationship, we determined median KJC values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median KJC ( K'JC(med) ). The obtained K'JC(med) values showed quite smaller deviation from the master curve at different temperatures than the experimental median KJC values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T0 value.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  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. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  1. 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.

  2. 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.

  3. 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.

  4. Slug impact loading on the vessel head during a postulated in-vessel steam explosion in pressurized water reactors -- Assessments and discussion of the investigation strategy

    SciTech Connect

    Krieg, R.; Malmberg, T.; Messemer, G.; Stach, T.; Stratmanns, E.

    1995-09-01

    The most severe consequence of a pressurized water reactor in-vessel steam explosion is a molten fuel slug impact against the head of the reactor pressure vessel that could cause a failure of this head and lead to missiles endangering the reactor containment. An investigation is described that attempts to determine the maximum slug impact that a vessel head is capable of withstanding without failing and, consequently, without impairing the containment safety-related function. Preliminary theoretical assessments are presented that suggest that the head might be able to withstand rather strong impacts and that the shape of the fuel slug will have only a moderate influence on the results, provided the upper internal structures are taken into account. A low sensitivity against the slug shape is an essential prerequisite for a reliable safety proof. However, investigations primarily based on computational models are not sufficient; therefore, an investigation concept is proposed that relies on model experiments in which the geometry is scaled down by factors of 10 and 20, respectively. Theoretical and experimental investigations for liquid-structure impact problems in different scales are discussed to assess the degree of similarity that can be obtained. Finally, model experiments are described in some detail simulating the molten fuel slug impact on the vessel head.

  5. 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.

  6. Co-gasification of different rank coals with biomass and petroleum coke in a high-pressure reactor for H(2)-rich gas production.

    PubMed

    Fermoso, J; Arias, B; Gil, M V; Plaza, M G; Pevida, C; Pis, J J; Rubiera, F

    2010-05-01

    Four coals of different rank were gasified, using a steam/oxygen mixture as gasifying agent, at atmospheric and elevated pressure in a fixed bed reactor fitted with a solids feeding system in continuous mode. Independently of coal rank, an increase in gasification pressure led to a decrease in H(2) + CO production and carbon conversion. Gasification of the different rank coals revealed that the higher the carbon content and reactivity, the greater the hydrogen production. Co-gasification experiments of binary (coal-biomass) and ternary blends (coal-petcoke-biomass) were conducted at high pressure to study possible synergetic effects. Interactions between the blend components were found to modify the gas production. An improvement in hydrogen production and cold gas efficiency was achieved when the coal was gasified with biomass.

  7. Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel

    DOE PAGES

    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

  8. 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

  9. 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

  10. HOx radical chemistry in oxidation flow reactors with low-pressure mercury lamps systematically examined by modeling

    DOE PAGES

    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

  11. 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

  12. 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".

  13. 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.

  14. 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.

  15. 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.

    PubMed

    Morfin, Franck; Piccolo, Laurent

    2013-09-01

    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(-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/TiO2 powder catalyst. PMID:24089839

  16. Influence of the reactor wall composition on radicals' densities and total pressure in Cl{sub 2} inductively coupled plasmas: I. Without silicon etching

    SciTech Connect

    Cunge, G.; Sadeghi, N.; Ramos, R.

    2007-11-01

    Laser absorption at 355 nm is used to monitor the time variations of the Cl{sub 2} density in high-density industrial inductively coupled plasma. This technique is combined with the measurement of the gas temperature from the Doppler width of the 811.5 nm line of argon, added as a trace gas and with the measurement of the total gas pressure with a Baratron gauge. These measurements permit to estimate the mole fractions of Cl{sub 2} and Cl species in Cl{sub 2} inductively coupled plasmas in a waferless reactor. The impact of the chemical nature of the reactor wall coatings on the Cl and Cl{sub 2} mole fractions is studied systematically. We show that under otherwise identical plasma conditions, the Cl mole fraction is completely different when the plasma is operated in SiOCl, AlF, CCl, or TiOCl coated reactors, because the homogeneous recombination probability of Cl atoms is strongly surface dependant. The Cl atom mole fraction reached at 100 W radiofrequency power in SiOCl coated reactor (80%) is much higher than that obtained at 900 W in a ''clean'' AlF reactor (40%). A simple zero-dimensional model permits to provide the recombination coefficient of Cl atoms, {gamma}{sub rec}: 0.005 on SiOCl film and about 0.3 on the other three coatings. It is proposed to get benefit of this very high sensitivity of Cl{sub 2} dissociation rate to the wall coating for the control of the chamber wall status from the Cl{sub 2} density measurements in standard conditions.

  17. Influence of the reactor wall composition on radicals' densities and total pressure in Cl2 inductively coupled plasmas: I. Without silicon etching

    NASA Astrophysics Data System (ADS)

    Cunge, G.; Sadeghi, N.; Ramos, R.

    2007-11-01

    Laser absorption at 355nm is used to monitor the time variations of the Cl2 density in high-density industrial inductively coupled plasma. This technique is combined with the measurement of the gas temperature from the Doppler width of the 811.5nm line of argon, added as a trace gas and with the measurement of the total gas pressure with a Baratron gauge. These measurements permit to estimate the mole fractions of Cl2 and Cl species in Cl2 inductively coupled plasmas in a waferless reactor. The impact of the chemical nature of the reactor wall coatings on the Cl and Cl2 mole fractions is studied systematically. We show that under otherwise identical plasma conditions, the Cl mole fraction is completely different when the plasma is operated in SiOCl, AlF, CCl, or TiOCl coated reactors, because the homogeneous recombination probability of Cl atoms is strongly surface dependant. The Cl atom mole fraction reached at 100W radiofrequency power in SiOCl coated reactor (80%) is much higher than that obtained at 900W in a "clean" AlF reactor (40%). A simple zero-dimensional model permits to provide the recombination coefficient of Cl atoms, γrec: 0.005 on SiOCl film and about 0.3 on the other three coatings. It is proposed to get benefit of this very high sensitivity of Cl2 dissociation rate to the wall coating for the control of the chamber wall status from the Cl2 density measurements in standard conditions.

  18. 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.

  19. Influence of the reactor wall composition on radicals' densities and total pressure in Cl{sub 2} inductively coupled plasmas: II. During silicon etching

    SciTech Connect

    Cunge, G.; Sadeghi, N.; Ramos, R.

    2007-11-01

    In an industrial inductively coupled plasma reactor dedicated to silicon etching in chlorine-based chemistry, the density of Cl{sub 2} molecules and the gas temperature are measured by means of laser absorption techniques, the density of SiCl{sub x} (x{<=}2) radicals by broadband absorption spectroscopy, the density of SiCl{sub 4} and ions by mass spectrometry, and the total gas pressure with a capacitance gauge. These measurements permit us to estimate the mole fractions of Cl, SiCl{sub 4}, and etch product radicals when etching a 200 mm diameter silicon wafer. The pure Cl{sub 2} plasma is operated in well prepared chamber wall coating with a thin film of SiOCl, AlF, CCl, or TiOCl. The impact of the chemical nature of the reactor wall's coatings on these mole fractions is studied systematically. We show that the reactor wall coatings have a huge influence on the radicals densities, but this is not only from the difference on Cl-Cl recombination coefficient on different surfaces. During silicon etching, SiCl{sub x} radicals sticking on the reactor walls are etched by Cl atoms and recycled into the plasma by forming volatile SiCl{sub 4}. Hence, the loss of Cl atoms in etching the wall deposited silicon is at least as important as their wall recombination in controlling the Cl atoms density. Furthermore, because SiCl{sub 4} is produced at high rate by both the wafer and reactor walls, it is the predominant etching product in the gas phase. However, the percentage of redeposited silicon that can be recycled into the plasma depends on the amount of oxygen present in the plasma: O atoms produced by etching the quartz roof window fix Si on the reactor walls by forming a SiOCl deposit. Hence, the higher the O density is, the lower the SiCl{sub 4} density will be, because silicon is pumped by the reactor walls and the SiOCl layer formed is not isotropically etched by chlorine. As a result, in the same pure Cl{sub 2} plasma at 20 mTorr, the SiCl{sub x} mole fraction can

  20. Influence of the reactor wall composition on radicals' densities and total pressure in Cl2 inductively coupled plasmas: II. During silicon etching

    NASA Astrophysics Data System (ADS)

    Cunge, G.; Sadeghi, N.; Ramos, R.

    2007-11-01

    In an industrial inductively coupled plasma reactor dedicated to silicon etching in chlorine-based chemistry, the density of Cl2 molecules and the gas temperature are measured by means of laser absorption techniques, the density of SiClx (x ⩽2) radicals by broadband absorption spectroscopy, the density of SiCl4 and ions by mass spectrometry, and the total gas pressure with a capacitance gauge. These measurements permit us to estimate the mole fractions of Cl, SiCl4, and etch product radicals when etching a 200mm diameter silicon wafer. The pure Cl2 plasma is operated in well prepared chamber wall coating with a thin film of SiOCl, AlF, CCl, or TiOCl. The impact of the chemical nature of the reactor wall's coatings on these mole fractions is studied systematically. We show that the reactor wall coatings have a huge influence on the radicals densities, but this is not only from the difference on Cl-Cl recombination coefficient on different surfaces. During silicon etching, SiClx radicals sticking on the reactor walls are etched by Cl atoms and recycled into the plasma by forming volatile SiCl4. Hence, the loss of Cl atoms in etching the wall deposited silicon is at least as important as their wall recombination in controlling the Cl atoms density. Furthermore, because SiCl4 is produced at high rate by both the wafer and reactor walls, it is the predominant etching product in the gas phase. However, the percentage of redeposited silicon that can be recycled into the plasma depends on the amount of oxygen present in the plasma: O atoms produced by etching the quartz roof window fix Si on the reactor walls by forming a SiOCl deposit. Hence, the higher the O density is, the lower the SiCl4 density will be, because silicon is pumped by the reactor walls and the SiOCl layer formed is not isotropically etched by chlorine. As a result, in the same pure Cl2 plasma at 20mTorr, the SiClx mole fraction can vary from 18% in a SiOCl-coated reactor, where the O density is the

  1. 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.

  2. 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. .

  3. 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…

  4. 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.

  5. Rapid pressure swing absorption cleanup of post-shift reactor synthesis gas. Technical progress report No. 4, June 1, 1991--September 31, 1991

    SciTech Connect

    Sirkar, K.K.; Majumdar, S.; Bhaumik, S.

    1991-10-31

    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. The specific objectives are (1) to fiber membranes. The specific objectives are (1) to develop a theoretical model which will provide guidelines for selecting an efficient RAPSAB process cycle which includes desorption; (2) to demonstrate the concept experimentally with a simple gas mixture (e.g., Co{sub 2}-N{sub 2}) and a simple absorbent liquid such as water, and (3) to extend the concept to reactive absorbent liquids for the separation of CO, Co{sub 2} from the post-shift reactor synthesis gas. A simplified theoretical description of the novel rapid pressure swing absorption process has been developed. The absorption part of the pressure swing absorption cycle has been predicted for CO{sub 2}-N{sub 2}-water system. Numerical simulation of the model is being carried out for different operating conditions for selecting an optimum pressure swing cycle.

  6. Reactor System Transient Code.

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  7. 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

  8. Extension Russian: A Few Notes

    ERIC Educational Resources Information Center

    Matthews, Irene J.

    1974-01-01

    Presents observations on the teaching of Russian in an evening extension course and four teaching techniques which involve variety in lesson planning, flexibility in presentation, a relaxed atmosphere and encouragement. (LG)

  9. 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...

  10. Astronauts Work in the Russian Zvezda Service Module

    NASA Technical Reports Server (NTRS)

    2001-01-01

    Astronauts Frank L. Culbertson, Jr. (left), Expedition Three mission commander, and Daniel W. Bursch, Expedition Four flight engineer, work in the Russian Zvezda Service Module on the International Space Station (ISS). Zvezda is linked to the Russian built Functional Cargo Block (FGB), or Zarya, the first component of the ISS. Zarya was launched on a Russian Proton rocket prior to the launch of Unity. The third component of the ISS, Zvezda (Russian word for star), the primary Russian contribution to the ISS, was launched by a three-stage Proton rocket on July 12, 2000. Zvezda serves as the cornerstone for early human habitation of the Station, providing living quarters, a life support system, electrical power distribution, a data processing system, a flight control system, and a propulsion system. It also provides a communications system that includes remote command capabilities from ground flight controllers. The 42,000 pound module measures 43 feet in length and has a wing span of 98 feet. Similar in layout to the core module of Russia's Mir space station, it contains 3 pressurized compartments and 13 windows that allow ultimate viewing of Earth and space.

  11. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  12. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  13. Novel, fiber optic, hybrid pressure and temperature sensor designed for high-temperature gen-IV reactor applications

    SciTech Connect

    Palmer, M. E.; Fielder, R. S.; Davis, M. A.

    2006-07-01

    A novel, fiber optic, hybrid pressure-temperature sensor is presented. The sensor is designed for reliable operation up to 1050 C, and is based on the high-temperature fiber optic sensors already demonstrated during previous work. The novelty of the sensors presented here lies in the fact that pressure and temperature are measured simultaneously with a single fiber and a single transducer. This hybrid approach will enable highly accurate active temperature compensation and sensor self-diagnostics not possible with other platforms. Hybrid pressure and temperature sensors were calibrated by varying both pressure and temperature. Implementing active temperature compensation resulted in a ten-fold reduction in the temperature-dependence of the pressure measurement. Sensors were also tested for operability in a relatively high neutron radiation environment up to 6.9x10{sup 17} n/cm{sup 2}. In addition to harsh environment survivability, fiber optic sensors offer a number of intrinsic advantages for nuclear power applications including small size, immunity to electromagnetic interference, self diagnostics / prognostics, and smart sensor capability. Deploying fiber optic sensors on future nuclear power plant designs would provide a substantial improvement in system health monitoring and safety instrumentation. Additional development is needed, however, before these advantages can be realized. This paper will highlight recent demonstrations of fiber optic sensors in environments relevant to emerging nuclear power plants. Successes and lessons learned will be highlighted. (authors)

  14. CATALisp - an expert system for real-time diagnosis of multiple-failure transients in pressurized water reactors

    SciTech Connect

    Martin, R.P.; Nassersharif, B. )

    1988-01-01

    The state of the art in artificial intelligence (AI) and expert system (ES) technology has matured to a degree that the potential development of a computer-aided/automated diagnostic and transient mitigation system in the area of nuclear reactor operation can be considered. Since traditional methods cannot handle complex systems efficiently, AI techniques provide a means to emulate an expert reactor operator rather than follow mechanistic methods. Computer-aided transient analysis coded in LISP (CATA-Lisp) is a confidence level based expert system written in Common LISP on the SYMBOLICS 3640 computer system. New versions are being developed in Common LISP for the Texas Instruments (TI) Explorer and the Sun microsystems machines, CATALisp manipulates both a knowledge base of transient identifier patterns (tree structured to allow for zooming in diagnostics) and a knowledge base containing a qualitative model of a nuclear power plant. The interference engine used by CATALisp uses the information stored in both knowledge bases to arrive at confidence level values that are used to infer particular plant states.

  15. Technology, safety and costs of decommissioning a reference pressurized water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    SciTech Connect

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies on conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference pressurized water reactor (PWR) described in the earlier study; defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs; and completing a study of recent PWR steam generator replacements to determine realistic estimates for time, costs and doses associated with steam generator removal during decommissioning. This report presents the results of recent PNL studies to provide supporting information in four areas concerning decommissioning of the reference PWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; assessing the cost and dose impacts of recent steam generator replacements; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  16. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  17. Time-dependent temper embrittlement of reactor pressure vessel steel: Correlation between microstructural evolution and mechanical properties during tempering at 650 °C

    NASA Astrophysics Data System (ADS)

    Li, Chuanwei; Han, Lizhan; Yan, Guanghua; Liu, Qingdong; Luo, Xiaomeng; Gu, Jianfeng

    2016-11-01

    The microstructural evolution of reactor pressure vessel (RPV) steel and its effect on the mechanical properties during tempering at 650 °C were studied to reveal the time-dependent toughness and temper embrittlement. The results show that the toughening of the material should be attributed to the decomposition of the martensite/austenite constituents and uniform distribution of carbides. When the tempering duration was 5 h, the strength of the investigated steel decreased to strike a balance with the material impact toughness that reached a plateau. As the tempering duration was further increased, the material strength was slightly reduced but the material impact toughness deteriorated drastically. This time-dependent temper embrittlement is different from traditional temper embrittlement, and it can be partly attributed to the softening of the matrix and the broadening of the ferrite laths. Moreover, the dimensions and distribution of the grain carbides are the most important factors of the impact toughness.

  18. Temper and neutron irradiation embrittlement in 2 1/4 Cr-1 Mo steels for pressure vessels of high-temperature gas-cooled reactors

    SciTech Connect

    Suzuki, M.; Fukaya, K.; Kodaira, T.; Oku, T.

    1984-09-01

    A 2 1/4 Cr-1 Mo steel is a promising candidate material for structural components of the pressure vessel of the experimental very high temperature gascooled reactor (VHTR) in Japan. Since the service temperature of such components is expected to be about 400/sup 0/C, the behavior of the temper and neutron irradiation embrittlements in these chromium-molybdenum steels should be confirmed from the viewpoint of structural integrity. The experimental verification on the degree of the embrittlement due to thermal aging, including the effect of applied stress and neutron irradiation, is described. Steel containing substantial amounts (about 100 ppm) of phosphorus atoms, which are believed to cause the temper embrittlement, showed that applied stress enhanced the embrittlement due to thermal aging. Embrittlement caused by neutron irradiation appears to be minimal in the case of the material containing <1000 ppm of copper as impurity with neutrons irradiated at about 400/sup 0/C.

  19. Evaluation of the recovery annealing of the reactor pressure vessel of NPP Nord (Greifswald) Units 1 and 2 by means of subsize impact specimens

    SciTech Connect

    Ahlstrand, R.; Klausnitzer, E.N.; Leitz, C.; Lange, D.; Pastor, D.; Valo, M.

    1993-12-01

    In 1988 and 1990, the reactor pressure vessels of Units 1 and 2, respectively, of the Greifswald nuclear power station were subjected to heat treatment at 475 C for annealing of irradiation effects. To demonstrate the effect of annealing and to evaluate a new postannealing transition temperature of vessel base metal and weld metal, boat samples were removed by means of electric discharge machining (EDM) from the (unclad) inner surface of the vessel. From these samples, micronotched bar impact test specimens were fabricated and tested at different temperature. Transition curves were evaluated from the results. By means of correlation tests, the transition temperatures evaluated from the micro-specimen tests are converted to standard Charpy-5 transition temperatures. Results are available for the weld metal of Unit 1 after annealing. The transition temperature T{sub k} is lower than the value calculated by the designer of the plant. Specimens removed from Unit 2 before and after annealing are in preparation.

  20. Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study

    SciTech Connect

    Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K.

    1998-03-01

    Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

  1. 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.

  2. Russian atomic energy reaches critical masses

    SciTech Connect

    Ustinov, A.

    1996-07-01

    With more than 200 malfunctions recorded in the early 1990`s the outmoded nuclear reactors in the former Soviet Union are in critical condition. Unfortunatly, there may not be enough money to monitor them, much less update them, says Artiom Ustinov, an expert in the Russian Ministry of Foreign Affairs in Moscow. In on short decade, Russia`s nuclear research budget has been slashed from $1.5 billion to $250 million. That amount of money simply cannot be stretched far enough to address safety issues, Ustinov says. {open_quotes}Currently, nuclear reactors in the former Soviet Union pose a greater environmental threat than ever before,{close_quotes} ustinov says. {open_quotes}In fact, almost half the nuclear power plants operating in Russia today are Chernobyl-like.{close_quotes} and most of these facilities are located in such large industrial centers as Moscow and st. Petersburg, Ustinov says. The future of nuclear power in the Soviet Union remains cloudy, Ustinov concludes. {open_quotes}It is likely that the legacy of its development will continue to generate electricity-and controversy-in the years ahead.{close_quotes}

  3. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  4. Propellant actuated nuclear reactor steam depressurization valve

    DOEpatents

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  5. 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)

  6. 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.

  7. 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.

  8. Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock

    SciTech Connect

    Sullivan, Edmund J.; Anderson, Michael T.

    2014-06-10

    This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

  9. Atmospheric pressure plasma chemical vapor deposition reactor for 100 mm wafers, optimized for minimum contamination at low gas flow rates

    SciTech Connect

    Anand, Venu E-mail: venuanand83@gmail.com; Shivashankar, S. A.; Nair, Aswathi R.; Mohan Rao, G.

    2015-08-31

    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.

  10. 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

  11. Dynamics of the atmospheric pressure diffuse dielectric barrier discharge between cylindrical electrodes in roll-to-roll PECVD reactor

    NASA Astrophysics Data System (ADS)

    Starostin, Sergey A.; Welzel, Stefan; Liu, Yaoge; van der Velden-Schuermans, Bernadette; Bouwstra, Jan B.; van de Sanden, Mauritius C. M.; de Vries, Hindrik W.

    2015-07-01

    The high current diffuse dielectric barrier discharge (DBD) was operated in a bi-axial cylindrical electrode configuration using nitrogen, oxygen and argon gas flow with the addition of tetraethyl orthosilicate as precursor for silica-like film deposition. The behaviour of the transient plasma was visualized by means of fast imaging from two orthogonal directions. The formation and propagation (~3 × 104 m s-1) of lateral ionization waves with the transverse light emission structure similar to the low pressure glow discharge was observed at time scales below 1 µs. Despite plasma non-uniformity at nanosecond time scale the deposition process on the web-rolled polymer results in smooth well adherent films with good film uniformity and excellent gas diffusion barrier properties. Contribution to the topical issue "The 14th International Symposium on High Pressure Low Temperature Plasma Chemistry (HAKONE XIV)", edited by Nicolas Gherardi, Ronny Brandenburg and Lars Stollenwark

  12. 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 ...

  13. Diagnostics of low-pressure hydrogen discharge created in a 13.56 MHz RF plasma reactor

    NASA Astrophysics Data System (ADS)

    Krištof, J.; Annušová, A.; Anguš, M.; Veis, P.; Yang, X.; Angot, T.; Roubin, P.; Cartry, G.

    2016-07-01

    A 13.56 MHz RF discharge in hydrogen was studied within the pressure range of 1-10 Pa, and at a power range of 400-1000 W. The electron energy distribution function and electron density were measured by a Langmuir probe. The gas temperature was determined by the Fulcher-α system in pure H2, and by the second positive system of nitrogen using N2 as the probing gas. The gas temperature was constant and equal to 450 ± 50 K in the capacitively coupled plasma (CCP) mode, and it increased with pressure and power in the inductively coupled plasma (ICP) mode. Also, the vibrational temperature of the ground state of hydrogen molecules was determined to be around 3100 and 2000 ± 500 K in the ICP and CCP mode, respectively. The concentration of atomic hydrogen was determined by means of actinometry, either by using Ar (5%) as the probing gas, or by using H2 as the actinometer in pure hydrogen (Q1 rotational line of Fulcher-α system). The concentration of hydrogen density increased with pressure in both modes, but with a dissociation degree slightly higher in the ICP mode (a factor 2).

  14. Diagnostics of low-pressure hydrogen discharge created in a 13.56 MHz RF plasma reactor

    NASA Astrophysics Data System (ADS)

    Krištof, J.; Annušová, A.; Anguš, M.; Veis, P.; Yang, X.; Angot, T.; Roubin, P.; Cartry, G.

    2016-07-01

    A 13.56 MHz RF discharge in hydrogen was studied within the pressure range of 1–10 Pa, and at a power range of 400–1000 W. The electron energy distribution function and electron density were measured by a Langmuir probe. The gas temperature was determined by the Fulcher-α system in pure H2, and by the second positive system of nitrogen using N2 as the probing gas. The gas temperature was constant and equal to 450 ± 50 K in the capacitively coupled plasma (CCP) mode, and it increased with pressure and power in the inductively coupled plasma (ICP) mode. Also, the vibrational temperature of the ground state of hydrogen molecules was determined to be around 3100 and 2000 ± 500 K in the ICP and CCP mode, respectively. The concentration of atomic hydrogen was determined by means of actinometry, either by using Ar (5%) as the probing gas, or by using H2 as the actinometer in pure hydrogen (Q1 rotational line of Fulcher-α system). The concentration of hydrogen density increased with pressure in both modes, but with a dissociation degree slightly higher in the ICP mode (a factor 2).

  15. 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.

  16. Effects of temperature, pressure, and carrier gas on the cracking of coal tar over a char-dolomite mixtures and calcined dolomite in a fixed-bed reactor

    SciTech Connect

    Seshadri, K.; Shamsi, A.

    1998-10-01

    A distillation fraction of a coal-derived liquid (tar) was cracked over a char-dolomite mixture, calcined dolomite, and silicon carbide in a fixed-bed reactor. The char-dolomite mixture (FWC) was produced from Pittsburgh No. 8 coal and dolomite in a Foster Wheeler carbonizer. The experiments were conducted under nitrogen and simulated coal gas (SCG), which was a mixture of CO, CO{sub 2}, H{sub 2}S, CH{sub 4}, N{sub 2}, and steam, at 1 and 17 atm. The conversion over these materials under nitrogen was much higher at 17 atm than at 1 atm. At higher pressures, tar molecules were trapped in the pores of the bed material and underwent secondary reactions, resulting in the formation of excess char. However, when nitrogen was replaced by SCG, the reactions that induce char formation were suppressed, thus increasing the yield of gaseous products. The analysis of the gaseous products and the spent bed materials for organic and inorganic carbons suggested that the product distribution can be altered by changing the carrier gas, temperature, and pressure.

  17. Synthesis and investigation of reaction mechanisms of diamondoids obtained by dielectric barrier discharge microplasma reactors operated in adamantane - argon - methane - hydrogen mixtures at atmospheric pressure

    NASA Astrophysics Data System (ADS)

    Stauss, Sven; Ishii, Chikako; Pai, David Z.; Terashima, Kazuo

    2013-09-01

    Diamondoids, sp3 hybridized molecules consisting of a cage-like carbon framework with hydrogen terminations, hold promise for many applications: biotechnology, medicine, and opto- and nanoelectronics. So far, diamondoids consisting of more than four cage units have been synthesized by electric discharge and pulsed laser plasmas in supercritical fluids, but the generation of plasmas in high-pressure media is not straightforward. Here we present an alternative, continuous flow process, where diamondoids are synthesized by dielectric barrier discharges inside microreactors. The plasmas were generated at peak-to-peak voltages of 3 - 4 kV at a frequency of 10 kHz, in Ar (96 - 100%-vol) - methane (0 - 4%-vol) - hydrogen (0 - 4%-vol) mixtures, at atmospheric pressure and flow rates of 2 - 20 sccm. As a precursor we used the first diamondoid, adamantane, whose density was controlled by adjusting the reactor temperature in the range from 293 to 323 K. Gas chromatography - mass spectrometry analysis indicated the synthesis of the second diamondoid, diamantane, and the presence of alkylated adamantane derivatives suggests a stepwise reaction mechanism. We will also discuss the influence of the plasma gas composition and precursor density on the diamondoid synthesis. Grant No. 21110002, MEXT, Japan.

  18. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

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

  19. 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…

  20. 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…

  1. 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…

  2. 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…

  3. 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.

  4. MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water

    NASA Astrophysics Data System (ADS)

    Bogoslovskaya, G. P.; Karpenko, A. A.; Kirillov, P. L.; Sorokin, A. P.

    2009-03-01

    Presented here is the MIF-SKD computation code for channel-by-channel thermohydraulic calculations of fuel-rod assemblies cooled by supercritical-pressure water. The code allows one to calculate distributions of temperature and velocity of the coolant in channels of fuel-rod assemblies and temperatures of fuel-rod claddings and fuel-rod assembly casings with regular and irregular geometrical characteristics at various distributions of energy release along the length and the radius of fuel-rod assemblies.

  5. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  6. A high-temperature, ambient-pressure ultra-dry operando reactor cell for Fourier-transform infrared spectroscopy.

    PubMed

    Köck, Eva-Maria; Kogler, Michaela; Pramsoler, Reinhold; Klötzer, Bernhard; Penner, Simon

    2014-08-01

    The construction of a newly designed high-temperature, high-pressure FT-IR reaction cell for ultra-dry in situ and operando operation is reported. The reaction cell itself as well as the sample holder is fully made of quartz glass, with no hot metal or ceramic parts in the vicinity of the high-temperature zone. Special emphasis was put on chemically absolute water-free and inert experimental conditions, which includes reaction cell and gas-feeding lines. Operation and spectroscopy up to 1273 K is possible, as well as pressures up to ambient conditions. The reaction cell exhibits a very easy and variable construction and can be adjusted to any available FT-IR spectrometer. Its particular strength lies in its possibility to access and study samples under very demanding experimental conditions. This includes studies at very high temperatures, e.g., for solid-oxide fuel cell research or studies where the water content of the reaction mixtures must be exactly adjusted. The latter includes all adsorption studies on oxide surfaces, where the hydroxylation degree is of paramount importance. The capability of the reaction cell will be demonstrated for two selected examples where information and in due course a correlation to other methods can only be achieved using the presented setup.

  7. A high-temperature, ambient-pressure ultra-dry operando reactor cell for Fourier-transform infrared spectroscopy

    SciTech Connect

    Köck, Eva-Maria; Kogler, Michaela; Pramsoler, Reinhold; Klötzer, Bernhard; Penner, Simon

    2014-08-15

    The construction of a newly designed high-temperature, high-pressure FT-IR reaction cell for ultra-dry in situ and operando operation is reported. The reaction cell itself as well as the sample holder is fully made of quartz glass, with no hot metal or ceramic parts in the vicinity of the high-temperature zone. Special emphasis was put on chemically absolute water-free and inert experimental conditions, which includes reaction cell and gas-feeding lines. Operation and spectroscopy up to 1273 K is possible, as well as pressures up to ambient conditions. The reaction cell exhibits a very easy and variable construction and can be adjusted to any available FT-IR spectrometer. Its particular strength lies in its possibility to access and study samples under very demanding experimental conditions. This includes studies at very high temperatures, e.g., for solid-oxide fuel cell research or studies where the water content of the reaction mixtures must be exactly adjusted. The latter includes all adsorption studies on oxide surfaces, where the hydroxylation degree is of paramount importance. The capability of the reaction cell will be demonstrated for two selected examples where information and in due course a correlation to other methods can only be achieved using the presented setup.

  8. Russian roulette with a knife.

    PubMed

    Frazer, M; Rosenberg, S

    1983-01-01

    The following case report describes an accidental stabbing that occurred on Halloween. The unwitting victim, while preparing for a holiday charade, stabbed himself with an ornamental dagger. By placing himself in a potentially hazardous situation, the victim's behavior entailed risk taking. The psychological implications of such risk taking are complex, and may be compared to such behaviors as russian roulette, gambling, and parasuicide.

  9. Atlanta Public Schools Russian Guide.

    ERIC Educational Resources Information Center

    Gaines, Billie Davis; And Others

    This guide for teachers of Russian outlines course objectives and general educational goals. Contents include information on: (1) philosophy and long-range goals, (2) student recruitment, (3) program counseling, (4) English in the classroom, (5) grammar, (6) articulation, (7) independent study, (8) grouping for student-centered work, (9) reading…

  10. Identity Options in Russian Textbooks

    ERIC Educational Resources Information Center

    Shardakova, Marya; Pavlenko, Aneta

    2004-01-01

    This article introduces a new analytical approach to the study of identity options offered in foreign and second language textbooks. This approach, grounded in poststructuralist theory and critical discourse analysis, is applied to 2 popular beginning Russian textbooks. Two sets of identity options are examined in the study: imagined learners…

  11. Russian program of planetary missions.

    PubMed

    Galeev, A A

    1996-01-01

    In the area of Solar System Exploration most of recently proposed mission oriented to the studies of Mars. Except MARS-96 and possibly MARS SAMPLE RETURN missions other Mars missions use Molnija class launchers. All Russian missions heavily involve international partners.

  12. 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.

  13. 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.

  14. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    NASA Astrophysics Data System (ADS)

    Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young

  15. Experience with detection and disposition of PWSCC flaws in PWR pressurizer and reactor coolant system loop Alloy 600 penetrations

    SciTech Connect

    Boggs, R.S.; Joseph, M.W.; Hall, J.F.

    1996-12-01

    The discovery of cracking of alloy 600 heater sleeve penetrations in a Combustion Engineering (CE) designed pressurizer in 1989, prompted utilities and the designer to form a group to investigate the cracking and address the potential safety concerns. The cause of cracking was determined to be axial primary water stress corrosion cracking (PWSCC). Its axial orientation made ejection of a penetration highly unlikely. Therefore, PWSCC of alloy 600 penetrations was not considered a safety issue but an economic issue. To manage the issue at the Florida Power and Light Co.`s (FPL`s) two unit St. Lucie Nuclear Power Plant, FPL has: performed visual inspections for evidence of PWSCC (leakage), preventively replaced penetrations based on predictions from industry data, and developed a standardized repair method as a contingency to respond to emergent repairs while minimizing the associated outage time.

  16. Decontamination of Aspergillus flavus and Aspergillus parasiticus spores on hazelnuts via atmospheric pressure fluidized bed plasma reactor.

    PubMed

    Dasan, Beyhan Gunaydin; Mutlu, Mehmet; Boyaci, Ismail Hakki

    2016-01-01

    In this study, an atmospheric pressure fluidized bed plasma (APFBP) system was designed and its decontamination effect on aflatoxigenic fungi (Aspergillus flavus and Aspergillus parasiticus) on the surface of hazelnuts was investigated. Hazelnuts were artificially contaminated with A. flavus and A. parasiticus and then were treated with dry air plasma for up to 5min in the APFBP system at various plasma parameters. Significant reductions of 4.50 log (cfu/g) in A. flavus and 4.19 log (cfu/g) in A. parasiticus were achieved after 5min treatments at 100% V - 25kHz (655W) by using dry air as the plasma forming gas. The decontamination effect of APFBP on A. flavus and A. parasiticus spores inoculated on hazelnuts was increased with the applied reference voltage and the frequency. No change or slight reductions were observed in A. flavus and A. parasiticus load during the storage of plasma treated hazelnuts whereas on the control samples fungi continued to grow under storage conditions (30days at 25°C). Temperature change on hazelnut surfaces in the range between 35 and 90°C was monitored with a thermal camera, and it was demonstrated that the temperature increase taking place during plasma treatment did not have a lethal effect on A. flavus and A. parasiticus spores. The damage caused by APFBP treatment on Aspergillus spp. spores was also observed by scanning electron microscopy.

  17. Decontamination of Aspergillus flavus and Aspergillus parasiticus spores on hazelnuts via atmospheric pressure fluidized bed plasma reactor.

    PubMed

    Dasan, Beyhan Gunaydin; Mutlu, Mehmet; Boyaci, Ismail Hakki

    2016-01-01

    In this study, an atmospheric pressure fluidized bed plasma (APFBP) system was designed and its decontamination effect on aflatoxigenic fungi (Aspergillus flavus and Aspergillus parasiticus) on the surface of hazelnuts was investigated. Hazelnuts were artificially contaminated with A. flavus and A. parasiticus and then were treated with dry air plasma for up to 5min in the APFBP system at various plasma parameters. Significant reductions of 4.50 log (cfu/g) in A. flavus and 4.19 log (cfu/g) in A. parasiticus were achieved after 5min treatments at 100% V - 25kHz (655W) by using dry air as the plasma forming gas. The decontamination effect of APFBP on A. flavus and A. parasiticus spores inoculated on hazelnuts was increased with the applied reference voltage and the frequency. No change or slight reductions were observed in A. flavus and A. parasiticus load during the storage of plasma treated hazelnuts whereas on the control samples fungi continued to grow under storage conditions (30days at 25°C). Temperature change on hazelnut surfaces in the range between 35 and 90°C was monitored with a thermal camera, and it was demonstrated that the temperature increase taking place during plasma treatment did not have a lethal effect on A. flavus and A. parasiticus spores. The damage caused by APFBP treatment on Aspergillus spp. spores was also observed by scanning electron microscopy. PMID:26398284

  18. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  19. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  20. Mobile Melt-Dilute Treatment for Russian Spent Nuclear Fuel

    SciTech Connect

    Peacock, H.

    2002-09-17

    Treatment of spent Russian fuel using a Melt-Dilute (MD) process is proposed to consolidate fuel assemblies into a form that is proliferation resistant and provides critically safety under storage and disposal configurations. Russian fuel elements contain a variety of fuel meat and cladding materials. The Melt-Dilute treatment process was initially developed for aluminum-based fuels so additional development is needed for several cladding and fuel meat combinations in the Russian fuel inventory (e.g. zirconium-clad, uranium-zirconium alloy fuel). A Mobile Melt-Dilute facility (MMD) is being proposed for treatment of spent fuels at reactor site storage locations in Russia; thereby, avoiding the costs of building separate treatment facilities at each site and avoiding shipment of enriched fuel assemblies over the road. The MMD facility concept is based on laboratory tests conducted at the Savannah River Technology Center (SRTC), and modular pilot-scale facilities constructed at the Savannah River Site for treatment of US spent fuel. SRTC laboratory tests have shown the feasibility of operating a Melt-Dilute treatment process with either a closed system or a filtered off-gas system. The proposed Mobile Melt-Dilute process is presented in this paper.

  1. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    SciTech Connect

    Jones, R.B.; Bolton, C.J.

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  2. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-09-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2-5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen.

  3. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    NASA Astrophysics Data System (ADS)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  4. High-temperature, high-pressure testing of zinc titanate in a bench-scale fluidized-bed reactor for 100 cycles

    SciTech Connect

    Gupta, R.P.; Gangwal, S.K.

    1993-06-01

    Integrated gasification combined cycle (IGCC) power plants are being advanced worldwide to produce electricity from coal owing to their potential for superior environmental performance, economics, and efficiency in comparison to conventional coal-based power plants. A key component of these plants is a hot-gas desulfurization system employing efficient regenerable mixed-metal oxide sorbents. Leading sorbent candidates include zinc ferrite and zinc titanate. These sorbents can remove hydrogen sulfide (H{sub 2}S) in the fuel gas down to very low levels (typically <20 ppmv) at 500 to 750{degree}C and can be readily regenerated for multicycle operation with air. To this end, the Research Triangle Institute (RTI) has formulated and tested a series of zinc titanate sorbents in a high-temperature, high- pressure HTHP fluidized-bed bench-scale reactor. Multicycle HTHP bench-scale testing of these sorbents under a variety of conditions culminated in the development of a ZT-4 sorbent that exhibited the best overall performance in terms of chemical reactivity, sulfur capacity, regenerability, structural properties, and attrition resistance. Following this parametric study, a life-cycle test consisting of 100 sulfidation-regeneration cycles was carried out with ZT-4 in the bench unit.

  5. Influence of stress intensity and loading mode on intergranular stress corrosion cracking of Alloy 600 in primary waters of pressurized water reactors

    SciTech Connect

    Rebak, R.B.; Szklarska-Smialowska, Z. . Fontana Corrosion Center)

    1994-05-01

    The steam generator in a pressurized water reactor (PWR) of a nuclear power plant consists mainly of a shell made of carbon (C) steel and tubes made of alloy 600 (UNS N06600). However, alloy 600 suffers environmentally induced cracking with exposure to high-temperature primary water. The susceptibility of alloy 600 to integranular stress corrosion cracking (IGSCC) was investigated as a function of the level of applied stresses and mode of loading. Constant load tests were conducted with specimens prepared from thin wall tubes, and constant deformation tests were conducted using specimens prepared from plates. With tubes exposed to primary water at 330 C, the crack propagation rate (CPR) was found to increase from 1 [times] 10[sup [minus]11] m/s at a stress intensity (K[sub i]) of 10 MPa[radical]m to 1 [times] 10[sup [minus]9] at K[sub i] = 60 MPa[radical]m. CPR obtained using compact specimens prepared from plates were 1 order of magnitude lower than values measured in tubes at the same temperature and in the same solution at each stress intensity. The corollary was that values of crack propagation and threshold stress intensities obtained using compact specimens could not be extrapolated to the behavior of thin wall tubes.

  6. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    PubMed

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  7. Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

  8. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  9. Russian low-level waste disposal program

    SciTech Connect

    Lehman, L.

    1993-03-01

    The strategy for disposal of low-level radioactive waste in Russia differs from that employed in the US. In Russia, there are separate authorities and facilities for wastes generated by nuclear power plants, defense wastes, and hospital/small generator/research wastes. The reactor wastes and the defense wastes are generally processed onsite and disposed of either onsite, or nearby. Treating these waste streams utilizes such volume reduction techniques as compaction and incineration. The Russians also employ methods such as bitumenization, cementation, and vitrification for waste treatment before burial. Shallow land trench burial is the most commonly used technique. Hospital and research waste is centrally regulated by the Moscow Council of Deputies. Plans are made in cooperation with the Ministry of Atomic Energy. Currently the former Soviet Union has a network of low-level disposal sites located near large cities. Fifteen disposal sites are located in the Federal Republic of Russia, six are in the Ukraine, and one is located in each of the remaining 13 republics. Like the US, each republic is in charge of management of the facilities within their borders. The sites are all similarly designed, being modeled after the RADON site near Moscow.

  10. Automatic safety rod for reactors

    DOEpatents

    Germer, John H.

    1988-01-01

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  11. Pressure surge attenuator

    DOEpatents

    Christie, Alan M.; Snyder, Kurt I.

    1985-01-01

    A pressure surge attenuation system for pipes having a fluted region opposite crushable metal foam. As adapted for nuclear reactor vessels and heads, crushable metal foam is disposed to attenuate pressure surges.

  12. The Modernization of Russian Education

    ERIC Educational Resources Information Center

    Filippov, V. M.

    2005-01-01

    These days many people are trying to assess what has been done in the past few years. The country is looking into the future and trying to map out the most optimal paths of its development. In this article, the author examines the main paths of development in the field of Russian Education. It is necessary to keep in mind the multi-plane character…

  13. Russian Arctic battles pipeline leak

    SciTech Connect

    Stone, R.

    1995-05-12

    An estimated 100,000 barrels of oil have leaked from a corroded pipeline in northern Russia`s Komi Republic since last August. The oil threatens salmon fisheries and marshlands used for grazing cattle and raindeer. This article discusses the spill in relation to other large spills and the possible ecological damage. Russian authorities, first reluctant to come to grips with the spill, are changing their attitudes, a position essential to gaining access to threatened regions for important monitoring studies.

  14. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  15. PRESSURE SYSTEM CONTROL

    DOEpatents

    Esselman, W.H.; Kaplan, G.M.

    1961-06-20

    The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

  16. The German/Russian MIR 1997 Mission: An Overview

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Session TP4 includes short reports concerning: (1) Life Science Experiments During the German-Russian Mir '97 Mission; (2) Orthostatic Intolerance Following Microgravity: A Role for Autonomic Dysfunction; (3) Heart Rate Variability and Skin Blood Flow in Man During Orthostatic Stress in Weightlessness; (4) Effects of Microgravity and Lower Body Negative Pressure on Circulatory Drives from Excercising Calf Muscles; (5) The Mir Station in Its Second Decade: Crew Science Operation During Mir '97; (6) Metabolic WARD (Water, Sodium, Calcium, and Bone Metabolism) and Endocrinological Experiments During the Mir '97 Mission; (7) Long-term Monitoring of the Spine-geometry During the Mir '97 Mission: Introduction of a New Method; and (8) Effects of 20 days of Microgravity (German/Russian Mir '97 Mission) on the Mechanical and Electromyographic Characteristics of Explosive Efforts of the Lower Limbs and of Cycloergometric Exercises of Mild to Sprint-Like Intensity.

  17. Neutron Dosimetry on the Full-Core First Generation VVER-440 Aimed at Reactor Support Structure Load Evaluation

    NASA Astrophysics Data System (ADS)

    Borodkin, P.; Borodkin, G.; Khrennikov, N.; Konheiser, J.; Noack, K.

    2009-08-01

    Reactor support structures (RSS), especially the ferritic steel wall of the water tank, of first-generation VVER-440 are non-restorable reactor equipment, and their lifetime may restrict plant-life. All operated Russian first generation VVER-440 have a reduced core with dummy assemblies except Unit 4 of Novovoronezh nuclear power plant (NPP). In comparison with other reactors, the full-core loading scheme of this reactor provides the highest neutron fluence on the reactor pressure vessel (RPV) and RSS accumulated over design service-life and its prolongation. The radiation load parameters on the RPV and RSS that have resulted from this core loading scheme should be evaluated by means of precise calculations and validated by ex-vessel neutron dosimetry to provide the reliable assessment of embrittlement parameters of these reactor components. The results of different types of calculations and their comparison with measured data have been analyzed in this paper. The calculational analysis of RSS fluence rate variation in dependence on the core loading scheme, including the standard and low leakage core as well as the introduction of dummy assemblies, is presented in this paper.

  18. Thermionic system evaluation test (TSET) facility construction: A United States and Russian effort

    SciTech Connect

    Wold, S.K. )

    1993-01-15

    The Thermionic System Evaluation Test (TSET) is a ground test of an unfueled Russian TOPAZ-II in-core thermionic space reactor powered by electric heaters. The facility that will be used for testing of the TOPAZ-II systems is located at the New Mexico Engineering Research Institute (NMERI) complex in Albuquerque, NM. The reassembly of the Russian test equipment is the responsibility of International Scientific Products (ISP), a San Jose, CA, company and Inertek, a Russian corporation, with support provided by engineers and technicians from Phillips Laboratory (PL), Sandia National Laboratories (SNL), Los Alamos National Laboratory (LANL), and the University of New Mexico (UNM). This test is the first test to be performed under the New Mexico Strategic Alliance agreement. This alliance consists of the PL, SNL, LANL, and UNM. The testing is being funded by the Strategic Defense Initiative Organization (SDIO) with the PL responsible for project execution.

  19. Thermionic system evaluation test (TSET) facility construction: A United States and Russian effort

    SciTech Connect

    Wold, S.K.

    1992-01-01

    The Thermionic System Evaluation Test (TSET) is a ground test of an unfueled Russian TOPAZ-II in-core thermionic space reactor powered by electric heaters. The facility that will be used for testing of the TOPAZ-II systems is located at the New Mexico Engineering Research Institute (NMERI) complex in Albuquerque, NM. The reassembly of the Russian test equipment is the responsibility of International Scientific Products (ISP), a San Jose, CA, company and Inertek, a Russian corporation, with support provided by engineers and technicians from Phillips Laboratory (PL), Sandia National Laboratories (SNL), Los Alamos National Laboratory (LANL), and the University of New Mexico (UNM). This test is the first test to be performed under the New Mexico Strategic Alliance agreement. This alliance consist of the PL, SNL, LANL, and UNM. The testing is being funded by the Strategic Defense Initiative Organization (SDIO) with the PL responsible for project execution.

  20. ISS Update: Russian Spacesuits and Spacewalks

    NASA Video Gallery

    NASA Public Affairs Officer Brandi Dean interviews Marc Ciupitu, EVA Flight and Increment Manager, about the Russian Orlan spacesuits that cosmonauts wear during spacewalks. Marc also discusses wha...