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Sample records for research reactor vessel

  1. 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.

  2. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  3. 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.

  4. 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.

  5. 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

  6. CERL/ORNL research and development programs in support of prestressed concrete reactor vessel development

    SciTech Connect

    Hornby, I.W.; Naus, D.J.

    1984-01-01

    In support of the evolution of PCRV designs being developed both in the UK and USA, research and developments programmers are being conducted at the CEGB Central Electricity Research Laboratories (CERL) and the Oak Ridge National Laboratory (ORNL) respectively. In the UK, recent work has focused on elevated temperature effects on concrete properties and instrument systems for PCRVs. The concrete development program at ORNL consists of generic studies designed to provide technical support for ongoing prestressed concrete reactor vessel-related activities, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities have been related to the development of properties for high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C lead plant project, and the development of PCRV model testing techniques.

  7. 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.

  8. 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.

  9. Prestressed-concrete reactor-vessel research and development studies at the Oak Ridge National Laboratory. [HTGR

    SciTech Connect

    Naus, D.J.

    1982-01-01

    The Prestressed Concrete Reactor Vessel (PCRV) research and development program at the Oak Ridge National Laboratory (ORNL) consists of generic studies to provide technical support for ongoing PCRV activities, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Current activities conducted in support of the development of high-temperature gas-cooled reactors (HTGRs) include: analysis methods development, concrete property determination, model testing, and component testing.

  10. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. 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.

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

    SciTech Connect

    Server, W. L.; Nanstad, Randy K

    2009-01-01

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

  13. Reactor vessel stud thread protector

    SciTech Connect

    Gasparro, M.R.

    1989-04-04

    This patent describes a stud thread protector for a nuclear reactor pressure vessel. The vessel has a removable closure head, the closure head being sealingly engaged with the pressure vessel by a plurality of stud bolts, an upper end thereof having a threaded section for threadingly engaging a nut and a vertical bore being disposed within the stud bolt. The stud thread protector encloses the exposed upper portion of the bolt and associated nut projecting above the closure head. The reactor vessel stud thread protector is comprised of: a tubular wall portion being opened at its lower end and substantially closed at its upper end; a drip pan associated with the outer surface of the protector, the drip pan being disposed radially inwardly with respect to the outer periphery of the vessel head, whereby the drip pan collects any fluid being emitted from the reactor vessel; and means for fastening the stud thread protector to an associated stud.

  14. Reactor vessel seal service fixture

    DOEpatents

    Ritz, W.C.

    1975-12-01

    An apparatus for the preparation of exposed sealing surfaces along the open rim of a nuclear reactor vessel comprised of a motorized mechanism for traveling along the rim and simultaneously brushing the exposed surfaces is described.

  15. Research Reactor MZFR, Karlsruhe, Germany Under Water Thermal Cutting of the Moderator Vessel and of the Thermal Shield

    SciTech Connect

    Loeb, A.; Eisenmann, B.; Prechtl, E.

    2006-07-01

    This paper presents the segmentation of the moderator vessel and of the thermal shield of the MZFR research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m{sup 3}. Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (authors)

  16. Reactor vessel annealing system

    DOEpatents

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  17. CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

    SciTech Connect

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01

    In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

  18. Reactor vessel lower head integrity

    SciTech Connect

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  19. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  20. Reactor vessel head permanent shield

    SciTech Connect

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-05-09

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs.

  1. 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.

  2. Nuclear reactor vessel fuel thermal insulating barrier

    SciTech Connect

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  3. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y.

    2012-07-01

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  4. Inspecting the reactor vessel penetrations

    SciTech Connect

    Bodson, F.; Fleming, K.W.

    1995-08-01

    The susceptibility of Alloy 600 to Primary Water Stress Corrosion Cracking (PWSCC) continues to plague nuclear power plants. Recently, the problem of PWSCC cracking has manifested itself in Control Rod Drive Mechanism (CRDM) head penetrations in nuclear plants in Europe. Framatome has been extensively involved in the performance of both inspections and repairs of CRDM head penetrations at Electricite de France (EdF) plants. B and W Nuclear Technologies (BWNT), building on Framatome technology, has developed a fully integrated service package and robotic manipulator to inspect and repair CRDM head penetrations for US utilities. Reactor vessel bottom penetration are also made of Alloy 600 and to tackle this potential PWSCC problem at EdF plants, Framatome has been performing specific inspections in order to detect the appearance of the phenomenon. This paper describes the overall range of inspection techniques and toolings developed to address these issues.

  5. 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.

  6. 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

  7. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  8. Neutron shielding panels for reactor pressure vessels

    SciTech Connect

    Singleton, Norman R.

    2011-11-22

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

  9. 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.

  10. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

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

    SciTech Connect

    Merriman, S.H.

    1980-10-01

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

  12. 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.

  13. 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.

  14. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  15. Limiting Factors for External Reactor Vessel Cooling

    SciTech Connect

    Cheung, F.B.

    2005-11-15

    The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

  16. TMI-2 reactor vessel head removal

    SciTech Connect

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  17. TMI-2 reactor vessel head removal

    SciTech Connect

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  18. 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.

  19. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

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

  20. BWR reactor vessel bottom head failure modes

    SciTech Connect

    Hodge, S.A.

    1989-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. The effect of the BWR procedural and structural differences upon the progression of a severe accident sequence during the period preceding movement of core debris into the reactor vessel lower plenum has been discussed previously. It is the purpose of this paper to briefly address the events occurring after debris relocation past the core plate and to describe the subsequent expected modes of bottom head pressure boundary failure. As an example, the calculated timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom Atomic Power Station is also presented. 14 refs., 4 figs., 1 tab.

  1. Reactor vessel fluence evaluation and dosimetry

    SciTech Connect

    Lois, L. )

    1992-01-01

    The methodology currently in use for the estimation of the fast neutron fluence to the pressure vessel (inside surface and reactor cavity) is based on discrete ordinates two-dimensional codes such as DOT or its updated version DORT. This methodology assumes a P[sub 3] scattering, an S[sub 8] quadrature approximation, and cross sections based on the ENDF/B-IV file. Associated one-dimensional codes are often used for the cross-section collapsing portion of the calculation. The neutron spectrum at the pressure vessel location of interest is estimated assuming a [sup 235]U, [sup 239]Pu, or [sup 241]Pu source spectrum or an appropriate combination thereof. The two-dimensional codes and associated methodologies were benchmarked in the early eighties using the results of the PCA and PSF Oak Ridge National Laboratory reactor experiments. The benchmarking experiments were estimated to provide an uncertainty of [approx]10%. The results of the calculations applied to a reactor were estimated to have an uncertainty of [approx]20%. This level of uncertainty was assumed in the estimation of the margin term defined in 10CFR50.61

  2. Reactor vessel cladding separate effects studies

    SciTech Connect

    Corwin, W.R.

    1985-01-01

    The existence of a layer of tough weld overlay cladding on the interior of a light-water reactor pressure vessel could mitigate damage caused during certain overcooling transients. The potential benefit of the cladding is that it could keep a short surface flaw, which would otherwise become long, from growing either by impeding crack initiation or by arresting a running crack. Two aspects critical to cladding behavior will be reported: irradiation effects on cladding toughness and the response of mechanically loaded, flawed structures in the presence of cladding. 15 refs., 24 figs., 6 tabs.

  3. 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.

  4. Welding for reactor vessel repair: Progress report

    SciTech Connect

    Birchenall, A.K.; Franco-Ferreira, E.A.

    1989-01-23

    Repair of intergranular stress corrosion cracking which may develop in SRP reactor vessels will be complicated by helium-induced weld cracking. The current leading candidate repair technique is a low heat input weld overlay which reduces the helium effect. Recent experiments show that low heat input Gas Metal Arc (GMA) weld overlays dramatically reduce helium-induced weld cracking in Type 304 stainless steel. The experiments indicate that helium-induced cracking is controlled by the heat input of the weld and the helium content of the material. Subsequent experiments may prove this type of weld overlay to be a suitable sealing technique for intergranular stress corrosion cracks (IGSCC) in irradiated stainless steel. 8 refs., 8 figs.

  5. Reactor pressure vessel stud management automation strategies

    SciTech Connect

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

    1992-01-01

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

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

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

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

  7. Dynamic analysis of large suspended LMFBR reactor vessels

    SciTech Connect

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1983-01-01

    Large breeder reactor vessels are often designed under the top-suspended condition. Since the vessel contains a large volume of liquid sodium as reactor coolant, the structural integrity of the vessel bottom head and its effect on the vessel dynamic response are of great importance to the safety and reliability of the reactor systems. This paper presents a dynamic analysis of the large suspended reactor vessel subjected to the horizontal earthquake excitation with the emphasis on the effect of bottom head vibration on fluid pressure and sloshing response. Unlike the conventional lumped mass method, the present analysis treats the liquid sodium as a continuum medium. As a result, the important effects ignored in the lumped mass method such as fluid coupling, fluid-structure interaction, interaction between sloshing and vessel vibration, etc. can be accounted into the analysis.

  8. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect

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

    2003-02-26

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

  9. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect

    Cooke, Conrad; Spann, Holger

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to

  10. A wall-crawling robot for reactor vessel inspection in advanced reactors

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected.

  11. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    NASA Astrophysics Data System (ADS)

    Gutsmiedl, Erwin; Scheuer, Anton

    2002-01-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256°C and 250°C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-impact-test samples were carried out in the temperature range between -256°C and 150°C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to be taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1×10 22 n/cm 2 was investigated. The loss of ductility was determined. As additional criteria the variation of the fracture toughness was studied. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfil the leak criteria before rupture

  12. PBF Reactor Building (PER620). Reactor vessel ready for insertion into ...

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

    PBF Reactor Building (PER-620). Reactor vessel ready for insertion into pit. Photographer: Holmes. Date: February 26, 1970. INEEL negative no. 70-991 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  13. PBF Reactor Building (PER620). Reactor vessel slips delicately into pit. ...

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

    PBF Reactor Building (PER-620). Reactor vessel slips delicately into pit. Photographer: Holmes. Date: February 26, 1970. INEEL negative no. 70-982 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  14. 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%).

  15. Reactor vessel using metal oxide ceramic membranes

    DOEpatents

    Anderson, Marc A.; Zeltner, Walter A.

    1992-08-11

    A reaction vessel for use in photoelectrochemical reactions includes as its reactive surface a metal oxide porous ceramic membrane of a catalytic metal such as titanium. The reaction vessel includes a light source and a counter electrode. A provision for applying an electrical bias between the membrane and the counter electrode permits the Fermi levels of potential reaction to be favored so that certain reactions may be favored in the vessel. The electrical biasing is also useful for the cleaning of the catalytic membrane.

  16. Investigation of vessel exterior air cooling for a HLMC reactor

    SciTech Connect

    Sienicki, J. J.; Spencer, B. W.

    2000-01-13

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  17. PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

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

    PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  18. Life extension of the St. Lucie unit 1 reactor vessel

    SciTech Connect

    Rowan, G.A.; Sun, J.B.; Mott, S.L. )

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program.

  19. Research Reactor Benchmarks

    SciTech Connect

    Ravnik, Matjaz; Jeraj, Robert

    2003-09-15

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given.

  20. PBF Reactor Building (PER620). Reactor vessel descends into pit, still ...

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

    PBF Reactor Building (PER-620). Reactor vessel descends into pit, still under control of handling beams and pulleys. Vertical-lift door (to Reactor Building) is in background. Photographer: Holmes. Date: February 26, 1970. INEEL negative no. 70-986 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  1. 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.

  2. The NRL-EPRI research program (RP886-2), evaluation and prediction of neutron embrittlement in reactor pressure vessel materials. Part 1: Dynamic C sub v, PCC sub v

    NASA Astrophysics Data System (ADS)

    Hawthorne, J. R.

    1980-12-01

    Nuclear reactor pressure vessel materials are subject to progressive reductions in fracture resistance in service due to neutron irradiation. Current technology is inadequate to quantitatively predict radiation embrittlement for all vessel materials and their metallurgical variations for the neutron fluences of interest. In addition, a relationship between apparent notch ductility and fracture toughness in the irradiated condition is needed to evolve more quantitative projections of structural integrity. The NRL-EPRI RP886-2 Program was formulated to advance both areas for the benefit of reactor vessel design and operation. Its primary objective is the development of a high quality data base for evaluation of current radiation embrittlement projection methods and the development of improved methods. This report documents program highlights and research results for CY 1979 along with plans for the completion of program investigations. Postirradiation test data are presented for plate, forging and weld deposit materials irradiated in six reactor experiments to fluences ranging from approx. 0.1 to approx. 10 to the 19th power n/sq cm = 1 MeV at 288 C. Comparisons are made between results for standard Charpy V-notch and fatigue precracked Charpy-V tests of preirradation and postirradiation material conditions. A companion document (Annual Progress Report for CY 1979: Part II) will present results for the 25.4 mm compact toughness (J-R curve) tests of the same materials and material conditions. A preliminary correlation of the Charpy-V and J-integral fracture toughness property changes with irradiation is observed.

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

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

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

  4. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  5. 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.

  6. Retrospective dosimetry analyses of reactor vessel cladding samples

    SciTech Connect

    Greenwood, L. R.; Soderquist, C. Z.; Fero, A. H.

    2011-07-01

    Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combined with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)

  7. PBF Reactor Building (PER620). Reactor vessel has been tilted upwards, ...

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

    PBF Reactor Building (PER-620). Reactor vessel has been tilted upwards, but handling device is too large to fit into pit. Workman uses torch to cut metal. Contrast the scale of the man vis-a-vis the (massive!) vessel. Photographer: Holmes. Date: February 26, 1970. INEEL negative no. 70-981 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  8. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  9. 46 CFR 3.05-3 - Oceanographic research vessel.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 1 2011-10-01 2011-10-01 false Oceanographic research vessel. 3.05-3 Section 3.05-3... OCEANOGRAPHIC RESEARCH VESSELS Definition of Terms Used in This Part § 3.05-3 Oceanographic research vessel. “An oceanographic research vessel is a vessel which the U.S. Coast Guard finds is employed exclusively in one...

  10. 46 CFR 3.05-3 - Oceanographic research vessel.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Oceanographic research vessel. 3.05-3 Section 3.05-3... OCEANOGRAPHIC RESEARCH VESSELS Definition of Terms Used in This Part § 3.05-3 Oceanographic research vessel. “An oceanographic research vessel is a vessel which the U.S. Coast Guard finds is employed exclusively in one...

  11. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  12. 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).

  13. Gaseous fuel reactor research

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  14. PBF Reactor Building (PER620). Reactor vessel arrives from gate city ...

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

    PBF Reactor Building (PER-620). Reactor vessel arrives from gate city steel at door of PBF. On flatbed, it is too high to fit under door. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-737 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  15. PBF Reactor Building (PER620). Bottom of reactor vessel shows beyond ...

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

    PBF Reactor Building (PER-620). Bottom of reactor vessel shows beyond handling beams. Hole at right is opening for coolant pipe. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-989 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  16. PBF Reactor Building (PER620). Camera looks into reactor vessel. Control ...

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

    PBF Reactor Building (PER-620). Camera looks into reactor vessel. Control rods are positioned at outer perimeter; transient rods, at inner perimeter. Photographer: Larry Page. Date: November 2, 1972. INEEL negative no. 72-5266 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  17. 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.

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

    SciTech Connect

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

    1986-01-01

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

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

    SciTech Connect

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

    1985-12-01

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

  20. 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.

  1. Code System to Calculate Probability of Reactor Vessel Failure.

    Energy Science and Technology Software Center (ESTSC)

    2000-04-24

    Version: 00 VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickelmore » content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjusted on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition. VISA2 is an upgraded release from the original VISA program developed by U.S. Nuclear Regulatory Commission staff. Improvements include a treatment of cladding effects; a more general simulation of flaw size, shape and location; a simulation of inservice inspection; a revised simulation of the reference temperature of the nil-ductility transition; and treatment of vessels with multiple welds and initial flaws.« less

  2. Annealing the reactor vessel at the Palisades Plant

    SciTech Connect

    Fenech, R.A.

    1996-03-01

    In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999.

  3. ETRCF, TRA654, INTERIOR. TEST VESSEL (NOT REACTOR) INSIDE PIT. INL ...

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

    ETR-CF, TRA-654, INTERIOR. TEST VESSEL (NOT REACTOR) INSIDE PIT. INL NEGATIVE NO. HD24-2-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  4. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  5. Development of an Inspection System for the Reactor Vessel/Containment Vessel of the PRISM and SAFR Liquid Metal Reactors

    SciTech Connect

    1989-02-01

    The integrity of the reactor vessel is of utmost importance in both the PRISM and SAFR concepts. The reactor vessel operates at elevated temperatures and contains molten liquid sodium. To ensure safe operation of the reactor, a periodic, visual inspection of the walls of the containment vessel is required by ASME specifications. This inspection would be conducted during a time when the reactor is shut down for refueling or maintenance. Nuclear Systems Associates, Inc. (NSA) was issued a PRDA contract by the Department of Energy to design, develop, and test a Closed Circuit Television (CCTV) camera system. The purpose of the system is to inspect the welds and wall surface of the Reactor Vessel/Container Vessel for both the PRISM and SAFR type reactors. The system was designed to function at the reactor's normal shutdown temperature, and provide a clear indication of flaws in the wall's weld seams and any cracks that might develop. The project was performed in three phases. The first phase concentrated the efforts on producing a compact camera system with the required resolution, self -contained lighting, and remote control focus and viewing angle. The proposed camera was then tested in a vessel mock-up and found to perform to required specifications at room (cold) temperatures. Simulated flaws, cracks, and a sodium leak were observed with required clarity on both a commercial and blackened stainless steel surfaces. The camera was tested with a single clear glass dome, a single coated glass dome, and a dual-glass dome covering the camera lens and mirror. The second phase of the project was conducted in two parts. The first part involved testing the vessel mock-up at elevated temperatures to verify that the required temperatures can be obtained. The mock-up was constructed with imbedded heaters and both control and indicating thermocouples. Stable operating temperatures over 400°F were achieved. During the second part of this phase, the camera was inserted into the

  6. 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.

  7. Concept for Dismantling the Reactor Vessel and the Biological Shield of the Compact Sodium-Cooled Nuclear Reactor Facility (KNK)

    SciTech Connect

    Hillebrand, I.; Benkert, J.

    2002-02-27

    The Compact Sodium-cooled Nuclear Reactor Facility (KNK) was an experimental nuclear power plant of 20 MW electric power erected on the premises of the Karlsruhe Research Center. The plant was initially run as KNK I with a thermal core between 1971 and 1974 and then, between 1977 and 1991, with a fast core as the KNK II fast breeder plant. Under the decommissioning concept, the plant is to be decommissioned completely to green field conditions at the end of 2005 in ten steps, i.e. under the corresponding ten decommissioning permits. To this day, nine decommissioning permits have been issued, the first one in 1993 and the most recent one, number nine, in 2001. The decommissioning and demolition activities covered by decommissioning permits 1 to 7 have been completed. Under the 8th Decommissioning Permit, the components of the primary system and the rotating reactor top shield are to be removed by late 2001. Under the 9th Decommissioning Permit, the reactor vessel with its internals, the primary shield, and the biological shield are to be dismantled. The residual sodium volume in the reactor vessel was estimated to amount to approx. 30 l. The maximum Co-60 activation is on the order of 107-108 Bq/g; the maximum dose rate in the middle of the vessel was measured in April 1997 to be 55 Sv/h. The difficulty involved especially in dismantling KNK, on the one hand, is posed by the residual sodium in the plant, which determines the choice of neither wet nor thermical techniques to be used in disassembly. Another difficulty is caused by the depth of activation by fast neutrons, as a result of which not only the reactor vessel proper, but also the entire primary shield (60 cm of grey cast iron) and large parts of the biological shield must be disassembled and disposed of under remote control.

  8. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOEpatents

    Ekeroth, Douglas E.; Orr, Richard

    1993-01-01

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  9. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOEpatents

    Ekeroth, D.E.; Orr, R.

    1993-12-07

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  10. Fabrication Flaws in Reactor Pressure Vessel Repair Welds

    SciTech Connect

    Schuster, George J.; Doctor, Steven R.

    2007-12-01

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

  11. High aspect reactor vessel and method of use

    NASA Technical Reports Server (NTRS)

    Wolf, David A. (Inventor); Sams, Clarence F. (Inventor); Schwarz, Ray P. (Inventor)

    1992-01-01

    An improved bio-reactor vessel and system useful for carrying out mammalian cell growth in suspension in a culture media are presented. The main goal of the invention is to grow and maintain cells under a homogeneous distribution under acceptable biochemical environment of gas partial pressures and nutrient levels without introducing direct agitation mechanisms or associated disruptive mechanical forces. The culture chamber rotates to maintain an even distribution of cells in suspension and minimizes the length of a gas diffusion path. The culture chamber design is presented and discussed.

  12. ETR, TRA642. ETR REACTOR VESSEL ARRIVES FROM BOSTON TO THE ...

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

    ETR, TRA-642. ETR REACTOR VESSEL ARRIVES FROM BOSTON TO THE GANTRY CRANE AT CENTRAL FACILITIES AREA OF NRTS. BOTTOM OF VESSEL FACES CAMERA, HAS OVER THIRTY OPENINGS. INL NEGATIVE NO. 56-4055. R.G. Larsen, Photographer, 12/17/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

    Code of Federal Regulations, 2012 CFR

    2012-01-01

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

  14. A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity

    SciTech Connect

    Simonen, Fredric A.

    2001-05-31

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.

  15. Design criteria for prestressed concrete reactor vessels for high-temperature reactors

    SciTech Connect

    Elter, C.; Becker, G.

    1982-11-01

    For the design and construction of prestressed concrete reactor vessels, data on loading, construction materials, and safety factors are required. A description is given of the design conditions according to the current state of technology in the Federal Republic of Germany. Special consideration is given to the allowable stresses and an appropriate proposal for such stresses is suggested.

  16. 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.

  17. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    SciTech Connect

    GRIFFIN, PATRICK J.

    1999-09-14

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

  18. Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

    NASA Astrophysics Data System (ADS)

    McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

    2012-12-01

    Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

  19. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

    1997-01-01

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  20. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOEpatents

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

  1. 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.

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

    SciTech Connect

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

    2014-10-28

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

  3. Ex-vessel fuel characterization results in the reactor building

    SciTech Connect

    Kobayashi, R.; Distenfeld, C. H.; Ferguson, D. E.

    1988-01-01

    Forced circulation during and after the Three Mile Island Unit 2 (TMI-2) accident distribution reactor fuel into the systems and components in the reactor building (RB) and auxiliary and fuel handling building (AFHB). The majority of the distributed fuel remained in the reactor coolant system in the RB, and some fuel traveled to the makeup and purification and waste disposal systems in the AFHB. Efforts began in 1985 and continued through 1987 to determine the location and amounts of ex-vessel fuel. The fuel characterization measurements were performed using three methods: gamma spectroscopy, visual inspection, and sampling. Gamma spectroscopy involves the detection of gamma radiation from tracer isotopes /sup 144/Ce or /sup 154/Eu to calculate the fuel quantity. Gamma spectroscopy was used in closed systems that were not directly accessible. Visual inspection with miniature cameras was used to determine the volume of fuel debris in accessible areas. Sampling was used in conjunction with both methods to either determine or verify the isotopic ratio to be used in the fuel calculations.

  4. 46 CFR 3.05-3 - Oceanographic research vessel.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 1 2012-10-01 2012-10-01 false Oceanographic research vessel. 3.05-3 Section 3.05-3 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY PROCEDURES APPLICABLE TO THE PUBLIC DESIGNATION OF OCEANOGRAPHIC RESEARCH VESSELS Definition of Terms Used in This Part § 3.05-3 Oceanographic research vessel....

  5. Development of a removable closure for large cavities in high temperature reactor prestressed concrete reactor vessels

    SciTech Connect

    Speidel, S.R.; Plettenberg, W.H.

    1980-06-01

    In concrete reactor vessels of high temperature reactors many cavities with different diameters are arranged. The design of the closures for such cavities is an important problem in the detail engineering. A draft based on using prestressed concrete as a material is presented for closures with high diameter values. The influence of tendon forces, creep and shrinkage, pressure, etc, on stresses, stress distribution, and deformation are proved by a finite element analysis and the results are pointed out. A short outlook of further development is given. 12 refs.

  6. 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.

  7. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Reactor Vessel Material Surveillance Program Requirements H Appendix H to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. H Appendix H to Part 50—Reactor Vessel Material Surveillance Program Requirements I. Introduction II....

  8. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Reactor Vessel Material Surveillance Program Requirements H Appendix H to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. H Appendix H to Part 50—Reactor Vessel Material Surveillance Program Requirements I. Introduction II....

  9. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Reactor Vessel Material Surveillance Program Requirements H Appendix H to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. H Appendix H to Part 50—Reactor Vessel Material Surveillance Program Requirements I. Introduction II....

  10. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Reactor Vessel Material Surveillance Program Requirements H Appendix H to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. H Appendix H to Part 50—Reactor Vessel Material Surveillance Program Requirements I. Introduction II....

  11. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Reactor Vessel Material Surveillance Program Requirements H Appendix H to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. H Appendix H to Part 50—Reactor Vessel Material Surveillance Program Requirements I. Introduction II....

  12. Insights from Investigations of In-Vessel Retention for High Powered Reactors

    SciTech Connect

    Joy L. Rempe

    2005-10-01

    In a three-year U.S. - Korean International Nuclear Energy Research Initiative (INERI), state-of-the-art analytical tools and key U.S. and Korean experimental facilities were used to explore two options, enhanced ERVC performance and the use of internal core catchers, that have the potential to increase the margin for in-vessel retention (IVR) in high power reactors (up to 1500 MWe). This increased margin has the potential to improve plant economics (owing to reduced regulatory requirements) and increase public acceptance (owing to reduced plant risk). Although this program focused upon the Korean Advanced Power Reactor -- 1400 MWe (APR 1400) design, recommentations were developed so that they can easily be applied to a wide range of existing and advanced reactor designs. This paper summarizes new data gained for evaluating the margin associated with various options investigated in this program. Insights from analyses completed with this data are also highlighted.

  13. Remote reactor vessel inspections at Savannah River Plant

    SciTech Connect

    Lewis, W.I. III; Goodwin, B.W.; French, T.J.

    1988-01-01

    The Equipment Engineering Division of the Savannah River Laboratory has recently completed the development of a robotic inspection system for use in conducting full volumetric examinations of the site's reactor tanks. The development of the inspection equipment was necessary to obtain quantitative volumetric data on the structural integrity of the tanks such that service life could be evaluated. The development program included the design and fabrication of four, five-axis robotic inspection tools, six three-axis camera and lighting masts, video equipment, mobile control centers, complete mockup and testing facilities, and the integration of ultrasonic and eddy current data collection equipment. This paper gives a brief history of past reactor vessel inspections. The need, justification, and implementation of the development effort is then discussed. The various system components developed including the tools, controls, data acquisition systems, etc. are discussed along with the unique design constraints each had to meet and how each supports the overall inspection system. Implementation of the inspection system in the field and future development acres to increase the equipments versatility are also described. 2 refs., 9 figs.

  14. ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

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

    ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

    SciTech Connect

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

    2006-02-15

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

  16. ETR, TRA642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED ...

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

    ETR, TRA-642. ON BASEMENT FLOOR. REACTOR VESSEL WILL BE PLACED WITHIN THE INNER METAL FORM. WHEN CONCRETE IS POURED OUTSIDE THIS FORM, CONDUIT HOLES WILL BE PRESERVE SPACE THROUGH HOLES. INL NEGATIVE NO. 56-1507. Jack L. Anderson, Photographer, 5/8/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  17. Heat exchanger integrated into the main vessel of a molten combustible salt reactor

    SciTech Connect

    Blum, J.M.; Ventre, E.

    1980-01-29

    Heat exchanger is integrated into the main vessel of a molten combustible salt reactor comprising a reactor skirt containing the active core, a main vessel surrounding the reactor skirt, pumps and primary exchangers, an outer vessel which doubles the main vessel, a thermostatic coolant between the main and outer vessels maintaining the main vessel wall at a temperature below the melting temperature of a crust of salt which is inactive from a nuclear standpoint and which forms a coating of solid salt protecting the inner surface of said main vessel. The calories are extracted from the core by means of autonomous heat transfer modules each comprising a primary exchanger and a pump, whereby each module is suspended in the intermediate space between the main vessel and the reactor skirt and supported by a bearing surface whose base is located on a cooperating bearing surface provided around an opening made in the wall of a supporting ferrule fixed close to the bottom of the reactor skirt and over the entire circumference of the latter, said ferrule extending from the skirt to the vicinity of the main vessel in the solid protective salt crust.

  18. 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.

  19. 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.

  20. 46 CFR 188.10-53 - Oceanographic research vessel.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Oceanographic research vessel. 188.10-53 Section 188.10-53 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OCEANOGRAPHIC RESEARCH VESSELS GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 188.10-53 Oceanographic research...

  1. 46 CFR 188.10-53 - Oceanographic research vessel.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Oceanographic research vessel. 188.10-53 Section 188.10-53 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OCEANOGRAPHIC RESEARCH VESSELS GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 188.10-53 Oceanographic research...

  2. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    NASA Astrophysics Data System (ADS)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (<800 °C). With the upgrade and development of advanced power reactors, however, enhancing the nucleate boiling rate and its upper limit, Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub

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

    SciTech Connect

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

    2007-01-01

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

  4. 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)

  5. Research reactor fork users manual

    SciTech Connect

    Hsue, S.T.; Menlove, H.O.; Bosler, G.E.; Dye, H.R.; Walton, G.; Halbig, J.K.; Siebelist, R.

    1993-11-01

    This manual describes the design features and operating characteristics of the research reactor fork. The system includes an ion chamber for gross gamma-ray counting, fission chambers for neutron counting, and a collimated high-resolution spectroscopy system for gamma-ray measurements. The neutron and ion chamber measurements are designed to be made underwater in spent-fuel cooling ponds. The neutron and gamma-ray detectors have been designed with high efficiencies to accommodate the relatively low emission rates of neutrons and gamma rays from low-burnup, research-type reactor fuel. This manual presents the design, performance, and test results for the system.

  6. 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.

  7. Design and analysis of a prestressed concrete reactor vessel for a 2000 MWe super large-sized BWR

    SciTech Connect

    James, R.J.; Rashid, Y.R.; Gou, P.F.; Sawyer, C.D.; Tanaka, S.; Shirai, Y.; Mori, M.; Takekuro, I.

    1996-06-01

    The super large-sized natural circulation BWR (Boiling Water Reactor) which eliminates the need for a short-term ECCS (Emergency Core Cooling System) is conceived as a sophisticated approach to future LWR (Light Water Reactor) plant design, promising simplified system design with improved safety and economics. This reactor vessel design requires the technology of prestressed concrete reactor vessels (PCRVs), because the large reactor vessel with a sufficient water inventory to eliminate ECCS is beyond the current technology of steel reactor vessels. A large power output of 2,000 MWe was selected for the initial scoping study.

  8. International Research Reactor Decommissioning Project

    SciTech Connect

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  9. Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1

    SciTech Connect

    Dunn, F.E.

    1987-01-01

    The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.

  10. The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

    SciTech Connect

    Dickson, T.L.; Simonen, F.A.

    1992-05-01

    Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel. 10 refs.

  11. The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

    SciTech Connect

    Dickson, T.L. ); Simonen, F.A. )

    1992-01-01

    Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel. 10 refs.

  12. LPT. EBOR (TAN646) reactor vessel, flow distribution tank. Outlet nozzle ...

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

    LPT. EBOR (TAN-646) reactor vessel, flow distribution tank. Outlet nozzle on side of vessel will be connected to coolant duct. Photographer: Lowin. Date: January 20, 1965. INEEL negative no. 65-237 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  13. 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.

  14. Intercalibration of research survey vessels on Lake Erie

    USGS Publications Warehouse

    Tyson, J.T.; Johnson, T.B.; Knight, C.T.; Bur, M.T.

    2006-01-01

    Fish abundance indices obtained from annual research trawl surveys are an integral part of fisheries stock assessment and management in the Great Lakes. It is difficult, however, to administer trawl surveys using a single vessel-gear combination owing to the large size of these systems, the jurisdictional boundaries that bisect the Great Lakes, and changes in vessels as a result of fleet replacement. When trawl surveys are administered by multiple vessel-gear combinations, systematic error may be introduced in combining catch-per-unit-effort (CPUE) data across vessels. This bias is associated with relative differences in catchability among vessel-gear combinations. In Lake Erie, five different research vessels conduct seasonal trawl surveys in the western half of the lake. To eliminate this systematic bias, the Lake Erie agencies conducted a side-by-side trawling experiment in 2003 to develop correction factors for CPUE data associated with different vessel-gear combinations. Correcting for systematic bias in CPUE data should lead to more accurate and comparable estimates of species density and biomass. We estimated correction factors for the 10 most commonly collected species age-groups for each vessel during the experiment. Most of the correction factors (70%) ranged from 0.5 to 2.0, indicating that the systematic bias associated with different vessel-gear combinations was not large. Differences in CPUE were most evident for vessels using different sampling gears, although significant differences also existed for vessels using the same gears. These results suggest that standardizing gear is important for multiple-vessel surveys, but there will still be significant differences in catchability stemming from the vessel effects and agencies must correct for this. With standardized estimates of CPUE, the Lake Erie agencies will have the ability to directly compare and combine time series for species abundance. ?? Copyright by the American Fisheries Society 2006.

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

    SciTech Connect

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

    2010-06-07

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

  16. 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.

  17. Initial Evaluation of the Heat-Affected Zone, Local Embrittlement Phenomenon as it Applies to Nuclear Reactor Vessels

    SciTech Connect

    McCabe, D.E.

    1999-09-01

    The objective of this project was to determine if the local brittle zone (LBZ) problem, encountered in the testing of the heat-affected zone (HAZ) part of welds in offshore platform construction, can also be found in reactor pressure vessel (RPV) welds. Both structures have multipass welds and grain coarsening along the fusion line. Literature was obtained that described the metallurgical evidence and the type of research work performed on offshore structure welds.

  18. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1995-01-01

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  19. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  20. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    SciTech Connect

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

  1. 46 CFR 188.10-53 - Oceanographic research vessel.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... and other marine geophysical or geological surveys, atmospheric research, and biological research. ... 46 Shipping 7 2011-10-01 2011-10-01 false Oceanographic research vessel. 188.10-53 Section 188.10-53 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OCEANOGRAPHIC RESEARCH...

  2. 46 CFR 188.10-53 - Oceanographic research vessel.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... and other marine geophysical or geological surveys, atmospheric research, and biological research. ... 46 Shipping 7 2010-10-01 2010-10-01 false Oceanographic research vessel. 188.10-53 Section 188.10-53 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OCEANOGRAPHIC RESEARCH...

  3. 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.

  4. 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.

  5. Simplified analysis of PRISM RVACS (Reactor Vessel Auxiliary Cooling System) performance without liner spill-over

    SciTech Connect

    Van Tuyle, G.J.

    1990-01-01

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs.

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

    SciTech Connect

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

    2011-07-01

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

  7. Cold source moderator vessel development for the High Flux Isotope Reactor: Thermal-hydraulic studies

    SciTech Connect

    Williams, P.T.; Lucas, A.T.; Wendel, M.W.

    1998-07-01

    A project is underway at Oak Ridge National Laboratory (ORNL) to design, test, and install a cold neutron source facility in the High Flux Isotope Reactor (HFIR). This new cold source employs supercritical hydrogen at cryogenic temperatures both as the medium for neutron moderation and as the working fluid for removal of internally-generated nuclear heating. The competing design goals of minimizing moderator vessel mass and providing adequate structural integrity for the vessel motivated the requirement of detailed multidimensional thermal-hydraulic analyses of the moderator vessel as a critical design subtask. This paper provides a summary review of the HFIR cold source moderator vessel design and a description of the thermal-hydraulic studies that were carried out to support the vessel development.

  8. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    SciTech Connect

    Sweeney, F.J. ); Carroll, D.G. ); Chen, C. ); Crane, C.; Dalton, R. ); Taylor, J.R. ); Tosunoglu, S. )

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS.

  9. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    SciTech Connect

    Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

    1993-01-01

    One of the most important safety systems in General Electric`s (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS.

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

    NASA Astrophysics Data System (ADS)

    Watts, Daniel; Finkenstadt, Daniel

    2012-02-01

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

  11. In-place nuclear reactor vessel annealing demonstration project

    SciTech Connect

    Howell, D.

    1994-09-01

    The objective of this paper was a discussion of the proposed annealing demonstration project at the canceled Marble Hill-1 reactor. The discussion, which was a compilation of transparencies on the noted subject, included overall objectives, scope of work, staging of equipment, and analytical objectives. Current status, including funding was summarized.

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

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S

    2006-01-01

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

  13. 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.

  14. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  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

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

  16. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  17. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  18. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

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

  19. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

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

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

    SciTech Connect

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

    2011-07-01

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

  1. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    SciTech Connect

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  2. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    SciTech Connect

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel

  3. MAGNESIUM MONO POTASSIUM PHOSPHATE GROUT FOR P-REACTOR VESSEL IN-SITU DECOMISSIONING

    SciTech Connect

    Langton, C.; Stefanko, D.

    2011-01-05

    The objective of this report is to document laboratory testing of magnesium mono potassium phosphate grouts for P-Reactor vessel in-situ decommissioning. Magnesium mono potassium phosphate cement-based grout was identified as candidate material for filling (physically stabilizing) the 105-P Reactor vessel (RV) because it is less alkaline than portland cement-based grout (pH of about 12.4). A less alkaline material ({<=} 10.5) was desired to address a potential materials compatibility issue caused by corrosion of aluminum metal in highly alkaline environments such as that encountered in portland cement grouts. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere. Fresh and cured properties were measured for: (1) commercially blended magnesium mono potassium phosphate packaged grouts, (2) commercially available binders blended with inert fillers at SRNL, (3) grouts prepared from technical grade MgO and KH{sub 2}PO{sub 4} and inert fillers (quartz sands, Class F fly ash), and (4) Ceramicrete{reg_sign} magnesium mono potassium phosphate-based grouts prepared at Argonne National Laboratory. Boric acid was evaluated as a set retarder in the magnesium mono potassium phosphate mixes.

  4. Reactivity Transients in Nuclear Research Reactors

    Energy Science and Technology Software Center (ESTSC)

    2015-01-01

    Version 01 AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant.

  5. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  6. Specific effects in microwave chemistry explored through reactor vessel design, theory, and spectroscopy.

    PubMed

    Ashley, Bridgett; Lovingood, Derek D; Chiu, Yu-Che; Gao, Hanwei; Owens, Jeffery; Strouse, Geoffrey F

    2015-11-01

    Microwave chemistry has revolutionized synthetic methodology for the preparation of organics, pharmaceuticals, materials, and peptides. The enhanced reaction rates commonly observed in a microwave have led to wide speculation about the function of molecular microwave absorption and whether the absorption leads to microwave specific effects and enhanced molecular heating. The comparison of theoretical modeling, reactor vessel design, and dielectric spectroscopy allows the nuance of the interaction to be directly understood. The study clearly shows an unaltered silicon carbide vessel allows measurable microwave penetration and therefore, molecular absorption of the microwave photons by the reactants within the reaction vessel cannot be ignored when discussing the role of molecular heating in enhanced molecular reactivity for microwave synthesis. The results of the study yield an improved microwave reactor vessel design that eliminates microwave leakage into the reaction volume by incorporating a noble metal surface layer onto a silicon carbide reaction vessel. The systematic study provides the necessary theory and measurements to better inform the arguments in the field. PMID:26280744

  7. Utilisation of British University Research Reactors.

    ERIC Educational Resources Information Center

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  8. Protective interior wall and attach8ing means for a fusion reactor vacuum vessel

    DOEpatents

    Phelps, Richard D.; Upham, Gerald A.; Anderson, Paul M.

    1988-01-01

    An array of connected plates mounted on the inside wall of the vacuum vessel of a magnetic confinement reactor in order to provide a protective surface for energy deposition inside the vessel. All fasteners are concealed and protected beneath the plates, while the plates themselves share common mounting points. The entire array is installed with torqued nuts on threaded studs; provision also exists for thermal expansion by mounting each plate with two of its four mounts captured in an oversize grooved spool. A spool-washer mounting hardware allows one edge of a protective plate to be torqued while the other side remains loose, by simply inverting the spool-washer hardware.

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

    SciTech Connect

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

    2007-01-01

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

  10. 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.

  11. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2003-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-08-01

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

  14. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    SciTech Connect

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

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

    SciTech Connect

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

    1998-01-01

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

  16. 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.

  17. Microwave release ofpectin from orange peel albedo using a closed vessel reactor system

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Pectin was extracted from blood Moro orange in a closed vessel reactor heated with microwave irradiation. Time of heating was either 2 minutes at 110 °C or 210 minutes at 75 °C in pH range of 1.7 to 2.8. The run at 75 °C and a pH 1.7 with resistive heating was performed to simulate industrial proces...

  18. 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.

  19. 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.

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

    NASA Astrophysics Data System (ADS)

    Lambrecht, M.; Almazouzi, A.

    2009-03-01

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

  1. Analysis of dpa Rates in the HFIR Reactor Vessel using a Hybrid Monte Carlo/Deterministic Method

    NASA Astrophysics Data System (ADS)

    Risner, J. M.; Blakeman, E. D.

    2016-02-01

    The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiation transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 in. below to approximately 12 in. above the height of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC0500OR22725 with the US Department of Energy. The US Government retains and the publisher, by accepting the article for publication, acknowledges that the US Government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for the US Government purposes.

  2. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    SciTech Connect

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-07-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

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

    SciTech Connect

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

    1990-10-01

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

  4. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  5. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    NASA Astrophysics Data System (ADS)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  6. 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.

  7. Supply of enriched uranium for research reactors

    SciTech Connect

    Mueller, H.

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  8. N.S. Savannah Reactor Vessel Metal Extraction and Radiochemical Analysis

    SciTech Connect

    Ranellone, Richard; Bowen, John; Stouky, Jon; Wiegand, John

    2008-01-15

    In early 2006 a project was concluded to determine radioisotopic inventory and Curie content of the N.S. Savannah Reactor Pressure Vessel (RPV), Internals and Neutron Shield Tank (NST) by extracting metal samples and performing radiochemical analysis. The objective of this project was to determine if the RPV and internals could be removed, packaged, shipped and disposed as Class A radioactive waste without opening the RPV or conducting further sampling of the RPV/Internals. The N.S. Savannah is de-fueled and has been shut down for 37 years. The following conclusions can be drawn from this project: - Results are consistent with previous analyses and are based upon conservative methodology and assumptions. - Nuclide concentration for the N/S Savannah reactor pressure vessel and internals package are shown to be within Class A disposal limits when averaged over the entire volume of metal in the Reactor Pressure Vessel and internals. - Performance of N.S. Savannah's nuclear reactor was excellent. During normal operations, the reactor seldom operated above 80% of its rated power level, thereby minimizing thermal stresses on the fuel cladding. In addition, the fuel rods were not subjected to any accident or severe transient conditions that could result in cladding breeches with subsequent release of fission products and fuel particles to the primary coolant loop. The trace quantities of Cesium-137 observed in the primary loop water indicate that some pinhole penetrations of fuel rod cladding may have occurred during operations. Another source of Cesium-137 could be the presence of uranium fuel on the exterior of the fuel rod cladding (tramp uranium), a condition not uncommon in the N.S. Savannah fuel fabrication time frame. Fissioning of this 'tramp uranium' would cause the rapid release of chemically active Cesium-137 into the reactor coolant. However, the absence of other fission products (e.g., Strontium-90) as well as uranium and transuranic isotopes in the reactor

  9. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  10. Furfural-based polymers for the sealing of reactor vessels dumped in the Arctic Kara Sea

    SciTech Connect

    HEISER,J.H.; COWGILL,M.G.; SIVINTSEV,Y.V.; ALEXANDROV,V.P.; DYER,R.S.

    1996-10-07

    Between 1965 and 1988, 16 naval reactor vessels were dumped in the Arctic Kara Sea. Six of the vessels contained spent nuclear fuel that had been damaged during accidents. In addition, a container holding {approximately} 60% of the damaged fuel from the No. 2 reactor of the atomic icebreaker Lenin was dumped in 1967. Before dumping, the vessels were filled with a solidification agent, Conservant F, in order to prevent direct contact between the seawater and the fuel and other activated components, thereby reducing the potential for release of radionuclides into the environment. The key ingredient in Conservant F is furfural (furfuraldehyde). Other constituents vary, depending on specific property requirements, but include epoxy resin, mineral fillers, and hardening agents. In the liquid state (prior to polymerization) Conservant F is a low viscosity, homogeneous resin blend that provides long work times (6--9 hours). In the cured state, Conservant F provides resistance to water and radiation, has high adhesion properties, and results in minimal gas evolution. This paper discusses the properties of Conservant F in both its cured and uncured states and the potential performance of the waste packages containing spent nuclear fuel in the Arctic Kara Sea.

  11. 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.

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

    SciTech Connect

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

    1991-01-01

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

  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. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    NASA Astrophysics Data System (ADS)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  15. 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.

  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. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    SciTech Connect

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-04-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.

  18. 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).

  19. Research on plasma core reactors

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  20. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  1. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    SciTech Connect

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs.

  2. Development of inspection systems for alloy 600 nozzles of PWR reactor vessel

    SciTech Connect

    Unate, K.; Ideo, M.; Sanagawa, T.; Shirai, T.; Araki, Y.

    1995-08-01

    PWR reactor vessels have alloy 600 nozzles at top and bottom heads. The former are head penetration nozzles for CRDM, and the latter are bottom mounted instrumentation nozzles. The authors have developed inspection systems of two types for each nozzle to confirm the soundness. ECT and UT Techniques are employed for both systems. These systems are controlled remotely and enable to reduce radiation exposure, inspection time and number of inspectors. Based on the functional tests using full scale mockups, the reliabilities and effectiveness of both systems were confirmed.

  3. Design of the reactor vessel inspection robot for the advanced liquid metal reactor

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a mock non-hostile environment and shown to perform as expected, as detailed in this report.

  4. Fiberoptic in-vessel viewing system for the International Thermonuclear Experimental Reactor

    NASA Astrophysics Data System (ADS)

    Heikkinen, Veli; Aikio, Mauri; Keranen, Kimmo; Wang, Minqiang

    2002-07-01

    A viewing system was designed and a prototype realized for the in-vessel inspection of the International Thermonuclear Experimental Reactor. The viewing is based on the line scanning principle, and the system consists of ten identical units installed on top of the reactor at 36deg intervals. Each device contains a laser, beam steering mirrors, and viewing probe with insertion mechanics. The probe has an outside diameter of 150 mm and a length of 14 m. The illumination design applies frequency-doubled Nd: yttrium-aluminum-garnet lasers whose beams are guided through hermetically sealed windows into the vacuum vessel. The diffuser optics creates a vertically oriented light stripe onto the vessel surface that is viewed by the imaging optics, consisting of 16 modules altogether covering horizontal and vertical field-of-views of 2deg and 162deg. The optical images are transferred to charge coupled device cameras via coherent fiber arrays. The multifocus design uses stacked fiber rows whose ends are assembled into different axial positions. The viewing probes rotate at a constant angular speed of 1deg/s and pictures are taken at 0.01deg intervals. The complete picture of the vessel interior is generated in 6 min producing 5.8 x109 image pixels. The image processing and analysis of possible defects in the vessel surfaces are performed off-line after the viewing procedure. A full-scale prototype of the viewing probe was constructed to demonstrate the feasibility of the design. Its illumination optics utilizes a diffractive optics element that transforms the collimated input beam into a rectangular output lobe with uniform intensity. The prototype has horizontal and vertical imaging optics field-of-views of 2deg and 12deg. The test results showed that the prototype can take pictures of good quality applying a continuously rotating probe having an angular speed of 0.08deg/s. Under optimum conditions, the minimum resolvable feature size at a 3 m distance is smaller than 1 mm

  5. 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.

  6. 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.

  7. 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.

  8. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab.

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

    SciTech Connect

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

    1990-03-01

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

  10. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  11. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  12. Facility modernization Annular Core Research Reactor

    SciTech Connect

    Morris, F.M.; Luera, T.F.; McCrory, F.M.; Nelson, D.A.; Trowbridge, F.R.; Wold, S.A.

    1990-07-01

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  13. 77 FR 60042 - Safety Zone; Research Vessel SIKULIAQ Launch, Marinette, WI

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ... Security FR Federal Register NPRM Notice of Proposed Rulemaking A. Regulatory History and Information The... SECURITY Coast Guard 33 CFR Part 165 RIN 1625-AA00 Safety Zone; Research Vessel SIKULIAQ Launch, Marinette... vessels from a portion of Menominee River during the launching of the Research vessel SIKULIAQ, on...

  14. 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

  15. 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.

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

    SciTech Connect

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

    2015-10-21

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

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

    DOE PAGESBeta

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

    2015-10-21

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

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

    SciTech Connect

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

    1986-03-01

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

  19. Thermal annealing of the reactor pressure vessel NPP Unit 2 in Jaslovske Bohunice for its radiation embrittlement regeneration

    SciTech Connect

    Kupca, L.; Cepcek, S.

    1993-12-01

    The status of the preparation works for the thermal annealing operation at reactor pressure vessel (RPV) V-230-type Unit 2 in Jaslovske Bohunice planned for Autumn 1993 is presented in this paper. The producer of the RPV W-213 type, SKODA Works, will perform the thermal annealing operation and manufacture all equipment needed. During the planned shutdown for the refueling operation of this unit in September 1989, samples were prepared from base material (BM) and weld metal (WM) by means of special equipment used for the analysis of the chemical composition in the Nuclear Power Plants Research Institute (VUJE) laboratories. Results of the analysis of the irradiated samples and the hardness measurements of RPV material before and after annealing operation serves as the measure of radiation embrittlement recovery efficiency. Possible extension of the operation life of RPVs of WWER type by means of suitable provisions during normal operation before thermal annealing is also discussed.

  20. Fuel behavior comparison for a research reactor

    NASA Astrophysics Data System (ADS)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

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

    PubMed

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

    2004-08-01

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

  2. Trends in fusion reactor safety research

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Holland, D. F.; Piet, S. J.

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex, with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g., for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.

  3. BWRSAR (Boiling Water Reactor Severe Accident Response) calculations of reactor vessel debris pours for Peach Bottom short-term station blackout

    SciTech Connect

    Hodge, S.A.; Ott, L.J.

    1988-01-01

    This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented. 4 refs., 13 figs., 3 tabs.

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

    NASA Astrophysics Data System (ADS)

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

    2013-11-01

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

  5. A Non-Heating Experimental Study on the Two-Phase Natural Circulation through the Annular Gap between Reactor Vessel and Insulation under External Vessel Cooling

    SciTech Connect

    Ha, K.S.; Park, R.J.; Cho, Y.R.; Kim, S.B.; Kim, H.D.; Kim, H.M.; Kim, K.Y.

    2004-07-01

    To improve the margin for IVR in high-power reactors, some design improvements of the vessel/insulation configuration to increase the heat removal rate by two-phase natural circulation have been proposed. To observe and evaluate the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulation in the APR1400 under external reactor vessel cooling, the T-HERMES program has been performed, that is, the THERMES- SCALE, T-HERMES-SMALL, HERMES-HALF, and T-HERMES-CFD studies. In this paper, the HERMES-HALF study, which is one of the T-HERMES programs, is introduced. The HERMES-HALF is a non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The objectives of this HERMES-HALF study are to observe and evaluate the two-phase natural circulation phenomena through the gap between the reactor vessel and the insulation in the APR1400. For these purposes, a half-scaled experimental facility is prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The behaviors of the boiling-induced two-phase natural circulation flow in the insulation gap are observed, and the liquid mass flow rates driven by the natural circulation loop and void fraction distribution are measured. And numerical analyses of the HERMES-HALF experiments using CFX-5.6 code have also been performed by solving unsteady, three-dimensional Reynolds-averaged Navier-Stokes equations for multiphase flows with the zero equation turbulence model. By the experimental flow observation and numerical predictions, weak recirculation flows in the near region of the shear key are observed. The void fraction monotonically increases from the water inlet to the shear key region. There exists a short decrease of the void fraction after passing through the shear key due to geometrical expansion and the recirculation flow caused by the shear key. The variation of

  6. 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

  7. Austrian contributions to reactor safety research

    NASA Astrophysics Data System (ADS)

    Sdouz, Gert

    1992-03-01

    An overview of Austrian contributions to reactor safety research is given. Starting with licensing for the power plant Zwenkendorf and participation in the safety projects PBF (Power Burst Facility) and LOFT (Loss Of Fluid Test), the work shifted later to participation in the study of international standard problems of the OECD and the IAEA. Since 1988 the center of activity is the calculation of the source term behavior of accident sequences, especially for reactors of Russian types. As an illustration the results of a source term calculation for a TMLB' sequence are presented.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. Tool for removing split pin remnants from a nuclear reactor vessel

    SciTech Connect

    Havoic-Conroy, S.M.

    1986-05-27

    A tool is described for removing from anchor bores in the upper core plate of a nuclear reactor vessel, anchor pin remnants which may be broken from guide tube assemblies upon withdrawal thereof from large seating apertures adjacent to the anchor bores in the upper core plate, the tool comprising: support means dimensioned to fit through a seating aperture, a pin driving member carried by the support means, means for lowering the support means through a seating aperture to a use position beneath the upper core plate with the pin driving member in registry with an adjacent anchor bore, and motive means for moving the pin driving member upwardly into the anchor bore to drive the anchor pin remnant therefrom.

  13. Structural dynamic and thermal stress analysis of nuclear reactor vessel support system

    NASA Technical Reports Server (NTRS)

    Chi-Diango, J.

    1972-01-01

    A nuclear reactor vessel is supported by a Z-ring and a box ring girder. The two proposed structural configurations to transmit the loads from the Z-ring and the box ring girder to the foundation are shown. The cantilever concrete ledge transmitting the load from the Z-ring and the box girder via the cavity wall to the foundation is shown, along with the loads being transmitted through one of the six steel columns. Both of these two supporting systems were analyzed by using rigid format 9 of NASTRAN for dynamic loads, and the thermal stresses were analyzed by AXISOL. The six column configuration was modeled by a combination of plate and bar elements, and the concrete cantilever ledge configuration was modeled by plate elements. Both configurations were found structurally satisfactory; however, nonstructural considerations favored the concrete cantilever ledge.

  14. 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.

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

    SciTech Connect

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

    1996-06-01

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

  16. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2014-04-01

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

  18. Examination of relocated fuel debris adjacent to the lower head of the TMI-2 reactor vessel

    SciTech Connect

    Akers, D.W.; Jensen, S.M.; Schuetz, B.K.

    1994-03-01

    As part of the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project, funded by the Organization for Economic Cooperation and Development, physical, metallurgical, and radiochemical examinations were performed on samples of previously molten material that had relocated to the lower plenum of the TMI-2 reactor during the accident of 28 March 1979. This report presents the results of those examinations and some limited analysis of these results as required for the interpretation of the data. Principal conclusions of the examinations are that the bulk lower head debris is homogeneous and composed primarily of (U,Zr)O{sub 2}. This molten material reached temperatures greater than 2,600 C and probably reached the lower head as a liquid or slurry at temperatures below the peak temperature. A debris bed was formed, which was composed of particular debris above a monolithic melt that solidified on the lower head.

  19. Method for coupled three-dimensional analysis of reactor vessel blowdowns with internal structures. [PWR

    SciTech Connect

    Silling, S.A.; Gross, M.B.; Santee, G.E. Jr.; Chang, F.H.

    1981-01-01

    The STEALTH 3D and WHAMSE 3D computer codes have been combined to perform three-dimensional coupled fluid/structure calculations of the blowdown response of a pressure vessel with internal structures typical of a pressurized water reactor. The fluid/structure coupling, which is performed cycle by cycle during a calculation, is described. The coupled fluid/structure code, STEALTH/WHAMSE 3D, has been used to simulate the decompression of test V31.1 from the HDR blowdown test series. Calculations of fluid pressure, differential fluid pressure and hoop strain compare favorably with the experimental data from test V31.1. The computed peak axial stain compares less favorably with the experimental data, probably due to coarseness of the structural grid. 14 refs.

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

    SciTech Connect

    Dickson, T.L.

    1993-04-01

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

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

    SciTech Connect

    Dickson, T.L.

    1993-01-01

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

  2. 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

  3. Radionuclide release from research reactor spent fuel

    NASA Astrophysics Data System (ADS)

    Curtius, H.; Kaiser, G.; Müller, E.; Bosbach, D.

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO 2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO 2-fuel (LWR fuel, enrichment in 235U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Jülich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl 2-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl x-Al and U 3Si 2-Al) was studied in 400 mL MgCl 2-rich salt brine in the presence of Fe 2+ under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH) 3(s) and Eu(OH) 3(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu(IV) oxyhydroxides species due to radiolytic effects cannot

  4. State-of-the-art for liquid-level measurements applied to in-vessel coolant level for nuclear reactors

    SciTech Connect

    Anderson, R.L.

    1980-01-01

    The TMI-2 accident indicated that a direct indication of the liquid level in the reactor vessel would have told the operators that the core was being uncovered. This state-of-the-cost survey covered the following methods: heated thermocouple, differential pressure, ultrasonic, capacitance, microwave, time-domain reflectometry, and externally mounted radiation detectors. (DLC)

  5. Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

    SciTech Connect

    1985-03-01

    This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.

  6. Thermal Properties of Structural Materials Found in Light Water Reactor Vessels

    SciTech Connect

    J. E. Daw; J. L. Rempe; D. L. Knudson

    2009-11-01

    High temperature material property data for structural materials used in existing Light Water Reactors (LWRs) are limited. Often, extrapolated values recommended in the literature differ significantly. To reduce such uncertainties, new data for SA533 Grade B, Class 1 (SA533B1) low alloy steel, Stainless Steel 304 (SS304), and Inconel 600, found in Light Water Reactor (LWR) vessels and penetrations, were acquired and tested using material property systems available at the High Temperature Test Laboratory (HTTL) at the Idaho National Laboratory (INL). Properties measured include thermal expansion, specific heat capacity, and thermal diffusivity for temperatures up to 1200 oC. From these results, thermal conductivity and density were calculated. Results show that, in some cases, previously recommended values for these material differ significantly from measured values at high temperatures. This is especially true for SA533B1, as previous data do not account for the phase transformation of this material between 740 oC and 840 oC.

  7. 46 CFR 3.10-1 - Procedures for designating oceanographic research vessels.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Procedures for designating oceanographic research... TO THE PUBLIC DESIGNATION OF OCEANOGRAPHIC RESEARCH VESSELS Designation § 3.10-1 Procedures for designating oceanographic research vessels. (a) Upon written request by the owner, master, or agent of...

  8. 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.

  9. 78 FR 58575 - Review of Experiments for Research Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer...

  10. 77 FR 18254 - Certificate of Alternative Compliance for the Research Vessel R/V SIKULIAQ

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-27

    ... SECURITY Coast Guard Certificate of Alternative Compliance for the Research Vessel R/V SIKULIAQ AGENCY... Compliance was issued for the research vessel R/V SIKULIAQ as required by 33 U.S.C. 1605(c) and 33 CFR 81.18... to the docket, call Renee V. Wright, Program Manager, Docket Operations, telephone...