Powered by Deep Web Technologies
Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

2

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

3

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

4

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

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

Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

1996-05-01T23:59:59.000Z

5

Neutron shielding panels for reactor pressure vessels  

DOE Patents [OSTI]

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.

Singleton, Norman R. (Murrysville, PA)

2011-11-22T23:59:59.000Z

6

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

7

Reactor Pressure Vessel Head Packaging & Disposal  

SciTech Connect (OSTI)

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.

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

2003-02-26T23:59:59.000Z

8

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Broader source: Energy.gov (indexed) [DOE]

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that 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 operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

9

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect (OSTI)

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.

Wang, Jy-An John [ORNL

2010-08-01T23:59:59.000Z

10

Fabrication Flaws in Reactor Pressure Vessel Repair Welds  

SciTech Connect (OSTI)

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.

Schuster, George J.; Doctor, Steven R.

2007-12-01T23:59:59.000Z

11

ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance  

SciTech Connect (OSTI)

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.

GRIFFIN, PATRICK J.

1999-09-14T23:59:59.000Z

12

Reactor pressure vessel head vents and methods of using the same  

DOE Patents [OSTI]

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.

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

2014-10-28T23:59:59.000Z

13

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

SciTech Connect (OSTI)

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) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

Walter, Matthew [Structural Integrity Associates, Inc.; Yin, Shengjun [ORNL; Stevens, Gary [U.S. Nuclear Regulatory Commission; Sommerville, Daniel [Structural Integrity Associates, Inc.; Palm, Nathan [Westinghouse Electric Company, Cranberry Township, PA; Heinecke, Carol [Westinghouse Electric Company, Cranberry Township, PA

2012-01-01T23:59:59.000Z

14

Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel  

SciTech Connect (OSTI)

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)

Yoo, C.; Km, B.; Chang, K.; Leeand, S. [Korea Atomic Energy Research Inst., 150 Dukjin-dong, Yuseung-gu, Daejeon 305-353 (Korea, Republic of); Park, J. [Chungnam National Univ., 220 Gung-dong, Yuseung-gu, Daejeon 305-764 (Korea, Republic of)

2006-07-01T23:59:59.000Z

15

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR (pressurized-water-reactor) plants  

SciTech Connect (OSTI)

Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1988-01-01T23:59:59.000Z

16

A methodology for determining fabrication flaws in a reactor pressure vessel  

SciTech Connect (OSTI)

The Pacific Northwest National Laboratory (PNNL) conducted a program with the major objective of estimating the rate of occurrence of fabrication flaws in US light-water reactor pressure vessels (RPVs). In this study, RPV mate4rial was examined using the Synthetic Aperture Focusing Technique for Ultrasonic Testing (SAFT-UT) to detect and characterize flaws created during fabrication. The inspection data obtained in this program has been analyzed to address the rates of flaw occurrence.

Schuster, G.J.; Doctor, S.R.; Simonen, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

1996-06-01T23:59:59.000Z

17

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

SciTech Connect (OSTI)

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.

GJ Schuster, FA Simonen, SR Doctor

2008-04-01T23:59:59.000Z

18

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports  

SciTech Connect (OSTI)

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.

Lu, S.C.; Sommer, S.C.; Johnson, G.L. (Lawrence Livermore National Lab., CA (USA)); Lambert, H.E. (FTA Associates, Oakland, CA (USA))

1990-10-01T23:59:59.000Z

19

BWR In-Core Monitor Housing Replacement Under Dry Condition of Reactor Pressure Vessel  

SciTech Connect (OSTI)

A new method of In-Core Monitor Housing replacement has been successfully applied to Tokai Unit 2 (BWR with 1100 MWe) in April of 2001. It was designed to replace a housing under dry condition of reactor pressure vessel (RPV): this enabled the elimination of water filled-up and drained processes during the replacement procedure resulting in the reduction of implementation schedule. To realize the dry condition, the radiation shields were placed in the RPV and the hollow guide pipe (GP) was adopted to transfer the apparatuses from the top to the bottom work area. (authors)

Tatsuo Ishida; Shoji Yamamoto; Fujitoshi Eguchi [Japan Atomic Power Company (Japan); Motomasa Fuse; Kouichi Kurosawa; Sadato Shimizu; Minoru Masuda [Hitachi Ltd. (Japan); Shinya Fujii; Junji Tanaka [General Electric International Inc. (Japan); Jacobson, Bryce A. [General Electric Company (United States)

2002-07-01T23:59:59.000Z

20

Research and Development Roadmaps for Nondestructive Evaluation of Cables, Concrete, Reactor Pressure Vessels, and Piping Fatique  

SciTech Connect (OSTI)

To address these research needs, the MAaD Pathway supported a series of workshops in the summer of 2012 for the purpose of developing R&D roadmaps for enhancing the use of Nondestructive Evaluation (NDE) technologies and methodologies for detecting aging and degradation of materials and predicting the remaining useful life. The workshops were conducted to assess requirements and technical gaps related to applications of NDE for cables, concrete, reactor pressure vessels (RPV), and piping fatigue for extended reactor life. An overview of the outcomes of the workshops is presented here. Details of the workshop outcomes and proposed R&D also are available in the R&D roadmap documents cited in the bibliography and are available on the LWRS Program website (http://www.inl.gov/lwrs).

Clayton, Dwight A [ORNL] [ORNL; Bakhtiari, Sasan [Argonne National Laboratory (ANL)] [Argonne National Laboratory (ANL); Smith, Cyrus M [ORNL] [ORNL; Simmons, Kevin [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Coble, Jamie [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Brenchley, David [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL); Meyer, Ryan [Pacific Northwest National Laboratory (PNNL)] [Pacific Northwest National Laboratory (PNNL)

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

ORNL/TM-2012/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Broader source: Energy.gov (indexed) [DOE]

2/380 2/380 Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program September 2012 Prepared by Cyrus Smith Randy Nanstad Robert Odette Dwight Clayton Katie Matlack Pradeep Ramuhalli Glenn Light DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via the U.S. Department of Energy (DOE) Information Bridge. Web site http://www.osti.gov/bridge Reports produced before January 1, 1996, may be purchased by members of the public from the following source. National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900

22

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

SciTech Connect (OSTI)

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 concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

23

Reactor vessel support system  

DOE Patents [OSTI]

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.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

24

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

SciTech Connect (OSTI)

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.

Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T. [Sandia National Labs., Albuquerque, NM (United States)

1996-05-01T23:59:59.000Z

25

Modular Inspection System for a Complete IN-Service Examination of Nuclear Reactor Pressure Vessel, Including Beltline Region  

SciTech Connect (OSTI)

Final Report for a DOE Phase II Contract Describing the design and fabrication of a reactor inspection modular rover prototype for reactor vessel inspection.

David H. Bothell

2000-04-30T23:59:59.000Z

26

Effects of 50/degree/C surveillance and test reactor irradiations on ferritic pressure vessel steel embrittlement  

SciTech Connect (OSTI)

The results of surveillance tests on the High-Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory revealed that a greater than expected embrittlement had taken place after about 17.5 effective full-power years of operation and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed. A research program was undertaken that included irradiating specimens in the Oak Ridge Research Reactor. Specimens of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged-arc weld fabricated at ORNL to reproduce the vessel seam weld. The results of the surveillance program and the materials research program performed in support of the evaluation of the HFIR pressure vessel are presented and show the welds to be more radiation resistant than the A212B. Results of irradiated tensile and annealing experiments are described as well as a discussion of mechanisms which may be responsible for enhanced hardening at low damage rates. 20 refs., 22 figs., 5 tabs.

Nanstad, R.K.; Iskander, S.K.; Rowcliffe, A.F.; Corwin, W.R.; Odette, G.R.

1988-01-01T23:59:59.000Z

27

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 (OSTI)

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.

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.

1991-10-01T23:59:59.000Z

28

Three-dimensional discrete ordinates radiation transport calculations of neutron fluxes for beginning-of-cycle at several pressure vessel surveillance positions in the high flux isotope reactor  

SciTech Connect (OSTI)

The objective of this research was to determine improved thermal, epithermal, and fast fluxes and several responses at mechanical test surveillance location keys 2, 4, 5, and 7 of the pressure vessel of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) for the beginning of the fuel cycle. The purpose of the research was to provide essential flux data in support of radiation embrittlement studies of the pressure vessel shell and beam tubes at some of the important locations.

Pace, J.V. III; Slater, C.O.; Smith, M.S.

1993-11-01T23:59:59.000Z

29

Reactor vessel support system. [LMFBR  

DOE Patents [OSTI]

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.

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

1980-05-09T23:59:59.000Z

30

Effect of temperature on the fracture toughness in the nuclear reactor pressure vessel steel (SA508-3)  

SciTech Connect (OSTI)

The elastic-plastic fracture toughness J{sub IC} of the Nuclear Reactor Pressure Vessel Steel (SA508-3) which has high toughness was obtained at three temperatures (room temperature, {minus}20 C, 200 C) using a 1/2 CT specimen. Especially the two methods recommended in ASTM and JSME were compared. It was found that difficulty exists in obtaining J{sub IC} by ASTM R-curve method, while JSME R-curve method yielded good results. The stretched zone width method gave slightly larger J{sub IC} values than those by the R-curve method for SA508-3 steel and the blunting line was not affected by the test temperatures. The relation between SZW and J, SZW and J/E and SZW and J/{sigma}{sub ys} before initiation of a stable crack growth in the fracture toughness test at three temperatures is described.

Oh, S.W.; Lim, M.B. [Dong-A Univ., Pusan (Korea, Republic of); Yoon, H.K. [Dong-Eui Univ., Pusan (Korea, Republic of)

1994-12-31T23:59:59.000Z

31

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents [OSTI]

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 structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

32

High pressure storage vessel  

DOE Patents [OSTI]

Disclosed herein is a composite pressure vessel with a liner having a polar boss and a blind boss a shell is formed around the liner via one or more filament wrappings continuously disposed around at least a substantial portion of the liner assembly combined the liner and filament wrapping have a support profile. To reduce susceptible to rupture a locally disposed filament fiber is added.

Liu, Qiang

2013-08-27T23:59:59.000Z

33

1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section  

SciTech Connect (OSTI)

The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

1994-11-01T23:59:59.000Z

34

Reactor vessel annealing system  

DOE Patents [OSTI]

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).

Miller, Phillip E. (Greensburg, PA); Katz, Leonoard R. (Pittsburgh, PA); Nath, Raymond J. (Murrysville, PA); Blaushild, Ronald M. (Export, PA); Tatch, Michael D. (Randolph, NJ); Kordalski, Frank J. (White Oak, PA); Wykstra, Donald T. (Pittsburgh, PA); Kavalkovich, William M. (Monroeville, PA)

1991-01-01T23:59:59.000Z

35

Pressure Vessel Burst Program: Automated hazard analysis for pressure vessels  

SciTech Connect (OSTI)

The design, development, and use of a Windows based software tool, PVHAZARD, for pressure vessel hazard analysis is presented. The program draws on previous efforts in pressure vessel research and results of a Pressure Vessel Burst Test Study. Prior papers on the Pressure Vessel Burst Test Study have been presented to the ASME, AIAA, JANNAF, NASA Pressure Systems Seminar, and to a DOD Explosives Safety Board subcommittee meeting. Development and validation is described for simplified blast (overpressure/impulse) and fragment (velocity and travel distance) hazard models. The use of PVHAZARD in making structural damage and personnel injury estimates is discussed. Efforts in-progress are reviewed including the addition of two-dimensional and three-dimensional (2D and 3D) hydrodynamic code analyses to supplement the simplified models, and the ability to assess barrier designs for protection from fragmentation.

Langley, D.R. [Aerospace Corp., Kennedy Space Center, FL (United States); Chrostowski, J.D. [ACTA Inc., Torrance, CA (United States); Goldstein, S. [Aerospace Corp., El Segundo, CA (United States); Cain, M. [General Physics Corp., Titusville, FL (United States)

1996-12-31T23:59:59.000Z

36

RIS-M-2186 INTERPRETATIOM OF STRAIN HBASUREMEMTS ON NUCLEAR PRESSURE VESSELS  

E-Print Network [OSTI]

pressure loadings of the vessel seen to be the reason in other regions. INIS-descriptors; BUR TYPE REACTORS

37

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents [OSTI]

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.

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

2013-03-19T23:59:59.000Z

38

Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base  

SciTech Connect (OSTI)

The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs.

Wang, J.A.; Kam, F.B.K.

1998-02-01T23:59:59.000Z

39

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

SciTech Connect (OSTI)

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)

Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2012-07-01T23:59:59.000Z

40

Boron determinations in pressure vessel steels  

SciTech Connect (OSTI)

Several studies have suggested that low-energy neutrons contribute to reactor pressure vessel (PV) embrittlement through interactions with boron impurities in the steel. Until now, the available information on boron contents in pressure vessel steels has been based on nominal concentrations or estimates provided by the materials manufacturers. To help resolve the question of boron contribution to PV steel embrittlement, samples of 38 different PV steels were analyzed by high-sensitivity gas mass spectrometry for their helium and boron contents. The boron contents were determined by measuring the increase in helium content in each material as a result of additional thermal neutron exposure. The results of these analyses showed natural boron contents that ranged from 0.23 to 5.11 wt. ppm in the various alloys.

Oliver, B.M. [Rockwell International Corp., Canoga Park, CA (United States). Rocketdyne Div.; McElroy, W.N. [Consultants and Technology Services, Richland, WA (United States); Kellogg, L.S. [Battelle-Pacific Northwest Lab., Richland, WA (United States); Farrar, H. IV

1994-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Impact of an apparent radiation embrittlement rate on the life expectancy of PWR (pressurized-water-reactor) vessel supports  

SciTech Connect (OSTI)

Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ /minus/ 10/sup 9/ n/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all LWR vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed for one of the plants indicate best-estimate critical flaw size corresponding to 32 EFPY, of /approximately/0.4 in. It appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size. Thus, presumably such flaws would have to exist at the time of fabrication. 19 refs., 8 figs., 3 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

42

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

SciTech Connect (OSTI)

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 the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01T23:59:59.000Z

43

International Hydrogen Fuel and Pressure Vessel Forum  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27–29, 2010 in Beijing, China. High pressure...

44

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

SciTech Connect (OSTI)

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

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10T23:59:59.000Z

45

Lightweight bladder lined pressure vessels  

DOE Patents [OSTI]

A lightweight, low permeability liner for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using torispherical or near torispherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film seamed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life.

Mitlitsky, Fred (1125 Canton Ave., Livermore, CA 94550); Myers, Blake (4650 Almond Cir., Livermore, CA 94550); Magnotta, Frank (1206 Bacon Way, Lafayette, CA 94549)

1998-01-01T23:59:59.000Z

46

Lightweight bladder lined pressure vessels  

DOE Patents [OSTI]

A lightweight, low permeability liner is described for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using tori spherical or near tori spherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film sealed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life. 19 figs.

Mitlitsky, F.; Myers, B.; Magnotta, F.

1998-08-25T23:59:59.000Z

47

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Proceedings from the forum, which took...

48

Fast neutron fluxes in pressure vessels using Monte Carlo methods  

SciTech Connect (OSTI)

The objective of this project is to determine the feasibility of calculating the fast neutron flux in the pressure vessel of a pressurized water reactor by Monte Carlo methods. Neutron reactions reduce the ductility of the steel and thus limit the useful life of this important reactor component. This work was performed for Virginia Power (VEPCO). VIM is a continuous-energy Monte Carlo code which provides a versatile geometrical capability and a neutron physics data base closely representing the EDNF/B-IV data from which it was derived.

Edlund, M.C.; Thomas, J.R.

1986-01-01T23:59:59.000Z

49

Pressurized reactor system and a method of operating the same  

DOE Patents [OSTI]

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

Isaksson, Juhani M. (Karhula, FI)

1996-01-01T23:59:59.000Z

50

Pressurized reactor system and a method of operating the same  

DOE Patents [OSTI]

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

Isaksson, J.M.

1996-06-18T23:59:59.000Z

51

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect (OSTI)

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17T23:59:59.000Z

52

Beryllium pressure vessels for creep tests in magnetic fusion energy  

SciTech Connect (OSTI)

Beryllium has interesting applications in magnetic fusion experimental machines and future power-producing fusion reactors. Chief among the properties of beryllium that make these applications possible is its ability to act as a neutron multiplier, thereby increasing the tritium breeding ability of energy conversion blankets. Another property, the behavior of beryllium in a 14-MeV neutron environment, has not been fully investigated, nor has the creep behavior of beryllium been studied in an energetic neutron flux at thermodynamically interesting temperatures. This small beryllium pressure vessel could be charged with gas to test pressures around 3, 000 psi to produce stress in the metal of 15,000 to 20,000 psi. Such stress levels are typical of those that might be reached in fusion blanket applications of beryllium. After contacting R. Powell at HEDL about including some of the pressure vessels in future test programs, we sent one sample pressure vessel with a pressurizing tube attached (Fig. 1) for burst tests so the quality of the diffusion bond joints could be evaluated. The gas used was helium. Unfortunately, budget restrictions did not permit us to proceed in the creep test program. The purpose of this engineering note is to document the lessons learned to date, including photographs of the test pressure vessel that show the tooling necessary to satisfactorily produce the diffusion bonds. This document can serve as a starting point for those engineers who resume this task when funds become available.

Neef, W.S.

1990-07-20T23:59:59.000Z

53

The Symmetry Properties of the Flow in a Nuclear Reactor Vessel  

Science Journals Connector (OSTI)

The turbulent flow in a Pressurized Water Reactor vessel is modeled in a small scale experiment. Careful observations and flow control experiments, driven by considerations of symmetry, show that this flow of ...

Pierre Albaréde

1996-01-01T23:59:59.000Z

54

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

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.

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

1984-12-01T23:59:59.000Z

55

TMI-2 reactor vessel head removal  

SciTech Connect (OSTI)

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.

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

1985-09-01T23:59:59.000Z

56

Liquid metal systems development: reactor vessel support structure evaluation  

SciTech Connect (OSTI)

Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration.

McEdwards, J.A.

1981-01-01T23:59:59.000Z

57

Engineering Test Reactor (ETR) Vessel Relocated after 50 years.  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

58

Gamma ray-induced embrittlement of pressure vessel alloys  

SciTech Connect (OSTI)

High-energy gamma rays emitted from the core of a nuclear reactor produce displacement damage in the reactor pressure vessel (RPV). The contribution of gamma damage to RPV embrittlement has in the past been largely ignored. However, in certain reactor designs the gamma flux at the RPV is sufficiently large that its contribution to displacement damage can be substantial. For example, gamma rays have been implicated in the accelerated RPV embrittlement observed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. In the present study, mechanical property changes induced by 10-MeV electron irradiation of a model Fe alloy and an RPV alloy of interest to the HFIR were examined. Mini-tensile specimens were irradiated with high-energy electrons to reproduce damage characteristic of the Compton recoil-electrons induced by gamma bombardment. Substantial increases in yield and ultimate stress were observed in the alloys after irradiation to doses up to 5.3x10{sup {minus}3} dpa at temperatures ({approximately}50{degrees}C) characteristic of the HFIR pressure vessel. These measured increases were similar to those previously obtained following neutron irradiation, despite the highly disparate nature of the damage generated during electron and neutron irradiation.

Alexander, D.E.; Rehn, L.E. [Argonne National Lab., IL (United States); Farrell, K.; Stoller, R.E. [Oak Ridge National Lab., TN (United States)

1994-11-01T23:59:59.000Z

59

Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum  

Broader source: Energy.gov [DOE]

Agenda for the International Hydrogen Fuel and Pressure Vessel Forum held Sept. 27-29, 2010, in Beijing, China

60

Scaling analysis for a reactor vessel mixing test  

SciTech Connect (OSTI)

The Westinghouse AP600 advanced pressurized water reactor design uses a gravity-forced safety injection system with two nozzles in the reactor vessel downcomer. In the event of a severe overcooling transient such as a steam-line break, this system delivers boron to the core to offset positive reactivity introduced by the negative moderator defect. To determine if the system design is capable of successfully terminating this type of reactivity transient, a test of the system has been initiated. The test will utilize a 1:9 scale model of the reactor vessel and cold legs. The coolant will be modeled with air, while the safety injection fluid will be simulated with a dense gas. To determine the necessary parameters for this model, a scaling analysis was performed. The continuity, diffusion, and axial Navier-Stokes equations for the injected fluid were converted into dimensionless form. A Boussinesq formulation for turbulent viscosity was applied in these equations. This procedure identified the Richardson, mixing Reynolds, diffusion Fourier, and Euler numbers as dimensionless groups of interest. Order-of-magnitude evaluation was used to determine that the Richardson and mixing Reynolds numbers were the most significant parameters to match for a similar experiment.

Radcliff, T.D.; Parsons, J.R.; Johnson, W.S. (Univ. of Tennessee, Knoxville, TN (United States)); Ekeroth, D.E. (Westinghouse Electric Corp., Pittsburgh, PA (United States))

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Design and dimensioning of pressure vessel for a marine substation.  

E-Print Network [OSTI]

?? This thesis presents the mechanical design and dimensioning of a pressure vessel, which is to be used as housing for a marine substation in… (more)

Eriksson, Lars

2010-01-01T23:59:59.000Z

62

Assessment of typical BWR (boiling water reactor) vessel configurations and examination coverage  

SciTech Connect (OSTI)

Even though boiling water reactors (BWRs) are not susceptible to the kind of incident known as pressurized thermal shock that must be considered in the design and operation of pressurized water reactors, BWR reactor pressure vessels (RPVs) have experienced higher than expected embrittlement caused by fast neutron irradiation. This has required the vessel to be at a higher temperature than originally projected before the plant can be taken to power operation. In addition, many BWR plants have received exemption from the 10-year volumetric nondestructive evaluation (NDE) of the vessel as required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B PV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components,'' because NDE access is severely restricted. Since many RPV welds have not been examined after being placed in service and the potential for service-induced flaws exists, regulatory authorities are looking closely at examination relief requests. BWR reactor vessel examination coverage is typically limited by plant design. Most BWR plants were designed when inservice examination codes were in the early stages of development, and very little consideration was give to designing for NDE access. Consequently, there is restricted access for many areas of the RPV. Since an increase in examination requirements has been placed in ASME B PV Code Section XI in these areas, efforts have begun on a thorough analysis of the vessel weld volumes examined during inservice examination and an evaluation of possibility expanding the RPV examination coverage. Because of these concerns, an investigation of the accessibility of the reactor vessel for NDE was performed to define the present status and to determine the improvements in coverage that can be accomplished in the near future. 7 refs., 9 figs., 4 tabs.

Walker, S.M. (EPRI Nondestructive Evaluation Center, Charlotte, NC (USA)); Feige, E.J.; Ingamells, J.R. (Southwest Research Inst., San Antonio, TX (USA)); Calhoun, G.L.; Davis, J.; Kapoor, A. (Westinghouse Electric Corp., Pittsburgh, PA (USA))

1990-10-01T23:59:59.000Z

63

Pressurized fluidized bed reactor and a method of operating the same  

DOE Patents [OSTI]

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

Isaksson, J.

1996-02-20T23:59:59.000Z

64

Experiment Hazard Class 5.3 High Pressure Vessels  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

3 High Pressure Vessels 3 High Pressure Vessels Applicability This hazard classification applies to working with pressure vessels and systems. Other hazard classifications and associated controls may apply to experiments in this hazard class. Experiment Category Experiments involving previously reviewed hazard controls are catergorized as medium risk experiments. Experiments involving new equipment, processes or materials, or modified hazard control schemes are categorized as high risk experiments. Hazard Control Plan Verification Statements Engineered Controls - The establishment of applicable controls in accordance with the (American Society of Mechanical Engineers) ASME Boiler and Pressure Code, ASME B.31 Piping Code and applicable federal, state, and local codes. Verify vessel is stampled with ASME Code Symbol or allowable

65

Creep of A508/533 Pressure Vessel Steel  

SciTech Connect (OSTI)

ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are allowed by Code Case N-499-2 (now incorporated as an appendix to Section III Division 5 of the Code). This Code Case was developed with a rather sparse data set and focused primarily on rolled plate material (A533 specification). Confirmatory tests of creep behavior of both A508 and A533 are described here that are designed to extend the database in order to build higher confidence in ensuring the structural integrity of the VHTR RPV during off-normal conditions. A number of creep-rupture tests were carried out at temperatures above the 371°C (700°F) Code limit; longer term tests designed to evaluate minimum creep behavior are ongoing. A limited amount of rupture testing was also carried out on welded material. All of the rupture data from the current experiments is compared to historical values from the testing carried out to develop Code Case N-499-2. It is shown that the A508/533 basemetal tested here fits well with the rupture behavior reported from the historical testing. The presence of weldments significantly reduces the time to rupture. The primary purpose of this report is to summarize and record the experimental results in a single document.

Richard Wright

2014-08-01T23:59:59.000Z

66

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

SciTech Connect (OSTI)

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.

Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

1989-12-01T23:59:59.000Z

67

Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

FORUM AGENDA FORUM AGENDA U.S. Department of Energy and Tsinghua University International Hydrogen Fuel and Pressure Vessel Forum Tsinghua University Beijing, PRC September 27 - 29, 2010 The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010 in Beijing, China. High pressure vessel experts gathered to share lessons learned from CNG and hydrogen vehicle deployments, and to identify R&D needs to aid the global harmonization of regulations, codes and standards to enable the successful deployment of hydrogen and fuel cell technologies. Forum Objectives: * Address and share data and information on specific technical topics discussed at the workshop in

68

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

SciTech Connect (OSTI)

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.

Simonen, Fredric A.

2001-05-31T23:59:59.000Z

69

Tokamak reactor first wall  

DOE Patents [OSTI]

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

70

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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.

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

71

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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.

Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

1993-01-01T23:59:59.000Z

72

Lightweight pressure vessels and unitized regenerative fuel cells  

SciTech Connect (OSTI)

Energy storage systems have been designed using lightweight pressure vessels with unitized regenerative fuel cells (URFCs). The vessels provide a means of storing reactant gases required for URFCs; they use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide a permeation barrier. URFC systems have been designed for zero emission vehicles (ZEVs); they are cost competitive with primary FC powered vehicles that operate on H/air with capacitors or batteries for power peaking and regenerative braking. URFCs are capable of regenerative braking via electrolysis and power peaking using low volume/low pressure accumulated oxygen for supercharging the power stack. URFC ZEVs can be safely and rapidly (<5 min.) refueled using home electrolysis units. Reversible operation of cell membrane catalyst is feasible without significant degradation. Such systems would have a rechargeable specific energy > 400 Wh/kg.

Mitlitsky, F.; Myers, B.; Weisberg, A.H.

1996-09-06T23:59:59.000Z

73

GRR/Section 6-HI-e - Boiler Pressure Vessel Permit | Open Energy  

Open Energy Info (EERE)

GRR/Section 6-HI-e - Boiler Pressure Vessel Permit GRR/Section 6-HI-e - Boiler Pressure Vessel Permit < GRR Jump to: navigation, search GRR-logo.png GEOTHERMAL REGULATORY ROADMAP Roadmap Home Roadmap Help List of Sections Section 6-HI-e - Boiler Pressure Vessel Permit 06HIGBoilerPressureVesselPermit.pdf Click to View Fullscreen Contact Agencies Hawaii Department of Labor and Industrial Relations Occupational Safety and Health Division Regulations & Policies Boiler and Pressure Vessel Regulations Triggers None specified Click "Edit With Form" above to add content 06HIGBoilerPressureVesselPermit.pdf Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Error creating thumbnail: Page number not in range. Flowchart Narrative Boiler/Pressure Vessel Permit

74

Pressure vessels and piping codes and standards: Volume 2. PVP-Volume 339  

SciTech Connect (OSTI)

The role of Codes and Standards for pressure vessels and piping has increased significantly over the past decade. More and more, developments in Codes and Standards are accommodating the increasing sophistication of analysis methods, the need to address post-construction and operating plant issues, and the efficiencies that may be gained by focusing codes and standards on the areas that present the greatest risk. Codes and Standards for new construction also have had to accommodate greater challenges and more extreme environments imposed by more escalating requirements on piping and pressure vessel design and fabrication. This volume has focused on these challenges faced by Codes and Standards development. The topics in this volume include: (1) International Code Developments; (2) Seismic Developments in Codes and Standards; (3) Fabrication, Repairs, and Installation Issues Relating to Codes and Standards; (4) Application of Risk Based Criteria to In-Service Inspections; (5) Risk Based Codes and Standards; (6) The Code--Then and Now; (7) Reactor Water Fatigue: Fitness for Service; and (8) Two ASME Pressure Technology Code Issues: Post-Construction Codes and Metrication. Separate abstracts were prepared for most of the papers in this volume.

Esselman, T.C. [ed.] [Altran Corp., Boston, MA (United States); Balkey, K. [ed.] [Westinghouse Electric Corp., Pittsburgh, PA (United States); Chao, K.K.N. [ed.] [Consumers Power Co., Jackson, MI (United States); Gosselin, S. [ed.] [Electric Power Research Institute, Charlotte, NC (United States); Hollinger, G. [ed.] [Babcock and Wilcox, Barberton, OH (United States); Lubin, B.T. [ed.] [ABB Combustion Engineering, Windsor, CT (United States); Mohktarain, K. [ed.] [CB and I Technical Services, Plainfield, IL (United States); O`Donnell, W. [ed.] [O`Donnell Consulting Engineers, Inc., Pittsburgh, PA (United States); Rao, K.R. [ed.] [Entergy Operations, Inc, Jackson, MI (United States)

1996-12-01T23:59:59.000Z

75

Tritium issues in commercial pressurized water reactors  

SciTech Connect (OSTI)

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

Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

2008-07-15T23:59:59.000Z

76

High-pressure Storage Vessels for Hydrogen, Natural Gas and Hydrogen-Natural Gas Blends  

Broader source: Energy.gov [DOE]

These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 – 29, 2010, in Beijing, China.

77

High-pressure Storage Vessels for Hydrogen, Natural Gas andHydrogen...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Gas and Blends - Materials Testing and Design Requirements for Hydrogen Components and Tanks International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Hydrogen...

78

Technical Forum Participants at the International Hydrogen Fuel and Pressure Vessel Forum  

Broader source: Energy.gov [DOE]

Photo of the Technical Forum Participants at the International Hydrogen Fuel and Pressure Vessel Forum, which was held on September 27–29, 2010, in Beijing, China.

79

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

DOE Patents [OSTI]

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.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1995-01-01T23:59:59.000Z

80

U.S. DOE Hydrogen and Fuel Cell Activities: 2010 International Hydrogen Fuel and Pressure Vessel Forum  

Broader source: Energy.gov [DOE]

Presentation at the International Hydrogen Fuel and Pressure Vessel Forum on September 27–29, 2010, in Beijing, China.

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Fuel assembly transfer basket for pool type nuclear reactor vessels  

DOE Patents [OSTI]

A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

1991-01-01T23:59:59.000Z

82

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

83

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Gluntz, D.M.; Taft, W.E.

1994-12-20T23:59:59.000Z

84

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

SciTech Connect (OSTI)

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

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

1984-02-01T23:59:59.000Z

85

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 (OSTI)

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.

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

2005-01-01T23:59:59.000Z

86

Lightweight cryogenic-compatible pressure vessels for vehicular...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

insulator surrounding the inner pressure container in the evacuated space to inhibit heat transfer. Additionally, vacuum loss from fuel permeation is substantially inhibited...

87

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

Broader source: Energy.gov (indexed) [DOE]

of one or more NDE techniques that can assist in the determination of current RPV fracture toughness as well as in prediction of fracture toughness with further aging of the...

88

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

SciTech Connect (OSTI)

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.

Langton, C.; Stefanko, D.

2011-01-05T23:59:59.000Z

89

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

DOE Patents [OSTI]

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.

Phelps, Richard D. (Greeley, CO); Upham, Gerald A. (Valley Center, CA); Anderson, Paul M. (San Diego, CA)

1988-01-01T23:59:59.000Z

90

Forum Agenda: International Hydrogen Fuel and Pressure Vessel...  

Broader source: Energy.gov (indexed) [DOE]

workshop in Washington, D.C. including testing and certification of Type 3 and Type 4 tanks, pressure relief device testing and validation, tank inspection, and end-of-life...

91

Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels  

SciTech Connect (OSTI)

A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: • Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions. • Perform creep tests and characterize the mechanisms of creep fracture process. • Quantify how the microstructure degradation controls the creep strength of welded steel specimens. • Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds. • Develop a microstructure-based creep fracture model to estimate RPVs service life . • Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates. • Simulate damage evolution in creep specimens by FE analyses. • Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage. • Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength. • Develop a fracture model for the structural integrity of RPVs subjected to creep loads. • Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

2013-11-26T23:59:59.000Z

92

Proceedings of PVP2007 2007 ASME Pressure Vessels and Piping Division Conference  

E-Print Network [OSTI]

Proceedings of PVP2007 2007 ASME Pressure Vessels and Piping Division Conference July 22-26, 2007 either to fatigue or environmentally-assisted cracking exacerbated by residual stresses introduced during can have significant effects on the susceptibility of a material to degradation mechanisms

Cambridge, University of

93

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents [OSTI]

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

94

Structural integrity assessment of type 201LN stainless steel cryogenic pressure vessels  

SciTech Connect (OSTI)

The ASME Boiler and Pressure Vessel Code Committee approved the Code Case 2123 in 1992 which allows the use of Type 201LN stainless steel in the construction of ASME Section VIII, Division 1 and Division 2 pressure vessels for -320{degrees}F applications. Type 201LN stainless steel is a nitrogen strengthened modified version of ASTM A240, Type 201 stainless steel with a restricted chemistry. The Code allowable design stresses for Type 201LN for Division 1 vessels are approximately 27% higher than Type 304 stainless steel and equal to that of the 5 Ni and 9 Ni steels. This paper discusses the important features of the Code Case 2123 and the structural integrity assessment of Type 201LN stainless steel cryogenic vessels. Tensile, Charpy-V-notch and fracture properties have been obtained on several heats of this steel including weldments. A linear-elastic fracture mechanics analysis has been conducted to assess the expected fracture mode and the fracture-critical crack sizes. The results have been compared with Type 304 stainless steel, 5 Ni and 9 Ni steel vessels.

Rana, M.D.; Zawierucha, R. [Praxair, Inc., Tonawanda, NY (United States)

1995-12-01T23:59:59.000Z

95

An evaluation of life extension of the HFIR pressure vessel. Supplement 1  

SciTech Connect (OSTI)

Preliminary analyses were performed in 1994 to determine the remaining useful life of the HFIR pressure vessel. The estimated total permissible life was {approximately} 50 EFPY (100 MW). More recently, the analyses have been updated, including a more precise treatment of uncertainties in the calculation of the hydrostatic-proof-test conditions and also including the contribution of gammas to the radiation-induced reduction in fracture toughness. These and other refinements had essentially no effect on the predicted useful life of the vessel or on the specified hydrostatic proof-test conditions.

Cheverton, R.D.

1996-08-01T23:59:59.000Z

96

File:06HIGBoilerPressureVesselPermit.pdf | Open Energy Information  

Open Energy Info (EERE)

HIGBoilerPressureVesselPermit.pdf HIGBoilerPressureVesselPermit.pdf Jump to: navigation, search File File history File usage File:06HIGBoilerPressureVesselPermit.pdf Size of this preview: 463 × 599 pixels. Other resolution: 464 × 600 pixels. Full resolution ‎(1,275 × 1,650 pixels, file size: 47 KB, MIME type: application/pdf) File history Click on a date/time to view the file as it appeared at that time. Date/Time Thumbnail Dimensions User Comment current 09:08, 24 October 2012 Thumbnail for version as of 09:08, 24 October 2012 1,275 × 1,650 (47 KB) Dklein2012 (Talk | contribs) 12:32, 23 October 2012 Thumbnail for version as of 12:32, 23 October 2012 1,275 × 1,650 (47 KB) Dklein2012 (Talk | contribs) 16:30, 24 July 2012 Thumbnail for version as of 16:30, 24 July 2012 1,275 × 1,650 (44 KB) Alevine (Talk | contribs)

97

Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor  

E-Print Network [OSTI]

A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated ...

Hejzlar, P.

98

The DOS 1 neutron dosimetry experiment at the HB-4-A key 7 surveillance site on the HFIR pressure vessel  

SciTech Connect (OSTI)

A comprehensive neutron dosimetry experiment was made at one of the prime surveillance sites at the High Flux Isotope Reactor (HFIR) pressure vessel to aid radiation embrittlement studies of the vessel and to benchmark neutron transport calculations. The thermal neutron flux at the key 7, position 5 site was found, from measurements of radioactivation of four cobalt wires and four silver wires, to be 2.4 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}. The thermal flux derived from two helium accumulation monitors was 2.3 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The thermal flux estimated by neutron transport calculations was 3.7 {times} 10{sup 12} n{center_dot}m{sup {minus}2}s{sup {minus}1}. The fast flux, >1 MeV, determined from two nickel activation wires, was 1.5 {times} 10{sup 12} n{center_dot}m{sup {minus}2}{center_dot}s{sup {minus}1}, in keeping with values obtained earlier from stainless steel surveillance monitors and with a computed value of 1.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}{sup {minus}1}. The fast fluxes given by two reaction-product-type monitors, neptunium-237 and beryllium, were 2.6 {times} 10{sup 13} n{center_dot}m{sup {minus}2}{center_dot}s {sup {minus}1} and 2.2 {times} 10{sup 13} n{center_dot}m{sup {minus}2}s{sup {minus}1}, respectively. Follow-up experiments indicate that these latter high values of fast flux are reproducible but are false; they are due to the creation of greater levels of reaction products by photonuclear events induced by an exceptionally high ratio of gamma flux to fast neutron flux at the vessel.

Farrell, K.; Kam, F.B.; Baldwin, C.A. [and others

1994-01-01T23:59:59.000Z

99

Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants  

SciTech Connect (OSTI)

Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

100

Emulsion polymerization of ethylene-vinyl acetate-branched vinyl ester using a pressure reactor system.  

E-Print Network [OSTI]

??A new pressure reactor system was designed to synthesize a novel branched ester-ethylene-vinyl acetate (BEEVA) emulsion polymer. The reactor system was capable of handling pressure… (more)

Tan, Chee Boon

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Large passive pressure tube light water reactor with voided calandria  

SciTech Connect (OSTI)

A reactor concept has been developed that can survive loss-of-coolant accidents (LOCAs) without scram and without replenishing primary coolant inventory while maintaining safe temperature limits on the fuel and pressure tube. The proposed concept is a pressure tube reactor of similar design to Canada deuterium uranium reactors but differing in three key aspects. First, a solid silicon carbide-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low-pressure gas instead of heavy water moderator, and this normally voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation while during a LOCA or loss of heat sink accident, it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. The fuel elements can operate under post-critical-heat-flux conditions even at full power without exceeding fuel design limits. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Although light water is used as both coolant and moderator, the reactor exhibits a high degree of neutron thermalization and a large prompt neutron lifetime, similar to D{sub 2}O-moderated cores. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile and inherent stability against xenon spatial oscillations.

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

1996-02-01T23:59:59.000Z

102

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

SciTech Connect (OSTI)

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.

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

1988-01-01T23:59:59.000Z

103

Structural integrity assessment of carbon and low-alloy steel pressure vessels using a simplified fracture mechanics procedure  

SciTech Connect (OSTI)

This paper describes a simplified fracture analysis procedure which was developed by Pellini to quantify fracture critical-crack sizes and crack-arrest temperatures of carbon and low-alloy steel pressure vessels. Fracture analysis diagrams have been developed using the simplified analysis procedure for various grades of carbon and low-alloy steels used in the construction of ASME, Section VIII, Division 1 pressure vessels. Structural integrity assessments have been conducted from the analysis diagrams.

Rana, M.D. (Praxair Inc., Tonawanda, NY (United States). Research and Development Dept.)

1994-08-01T23:59:59.000Z

104

Integrity management of a HIC-damaged pipeline and refinery pressure vessel through hydrogen permeation measurements  

SciTech Connect (OSTI)

Hydrogen permeation measurements were used in the successful operation of a sour gas pipeline subsequent to a hydrogen-induced cracking (HIC) failure in September 1992. Two joints of HIC-resistant pipe were used to repair the failed section and adjacent cut-outs. The pipeline has been operated for five years with no further instances of HIC failure. Hydrogen permeation monitoring was chosen as an integrity management tool because no techniques are currently available to inspect for HIC damage in a pipeline this size. Self-powered electrochemical devices installed on the pipeline were employed to monitor and control the effectiveness of a batch inhibition program in maintaining diffusing hydrogen atom concentrations below the laboratory-measured threshold for initiation of HIC damage. Permeation monitoring of a HIC-damaged refinery pressure vessel indicated very high hydrogen atom flux, despite attempts to inhibit corrosion with ammonium polysulfide injection. In this instance it was decided that replacement of the vessel was necessary.

Hay, M.G.; Rider, D.W. [Shell Canada Ltd., Calgary, Alberta (Canada)

1998-12-31T23:59:59.000Z

105

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS  

SciTech Connect (OSTI)

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume 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 does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk will again be very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.32 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2009-12-29T23:59:59.000Z

106

Atmospheric pressure plasma processing with microstructure electrodes and microplanar reactors  

Science Journals Connector (OSTI)

Atmospheric pressure plasmas can be generated, if the distance between the plasma generating electrodes is in the range of 100 ?m, and radio-frequencies of 13.56 or 27.12 \\{MHz\\} are applied. Such small dimensioned plasmas are only of interest for industrial plasma applications if larger areas can be processed. It will be shown that both with microstructure electrodes as with microplanar-reactor, plasma processing can be carried out for typical substrate dimensions of 100 mm and more using helium or neon for plasma generation. First experiments of plasma surface treatment of polymers and of thin film deposition on silicon will be presented. With mixtures of some percentage C2H2 in atmospheric pressure helium, diamond-like carbon films with deposition rates between 1–10 ?m/min can be deposited.

H. Schlemm; D. Roth

2001-01-01T23:59:59.000Z

107

Manufacturing Cost Analysis of Novel Steel/Concrete Composite Vessel for Stationary Storage of High-Pressure Hydrogen  

SciTech Connect (OSTI)

A novel, low-cost, high-pressure, steel/concrete composite vessel (SCCV) technology for stationary storage of compressed gaseous hydrogen (CGH2) is currently under development at Oak Ridge National Laboratory (ORNL) sponsored by DOE s Fuel Cell Technologies (FCT) Program. The SCCV technology uses commodity materials including structural steels and concretes for achieving cost, durability and safety requirements. In particular, the hydrogen embrittlement of high-strength low-alloy steels, a major safety and durability issue for current industry-standard pressure vessel technology, is mitigated through the use of a unique layered steel shell structure. This report presents the cost analysis results of the novel SCCV technology. A high-fidelity cost analysis tool is developed, based on a detailed, bottom-up approach which takes into account the material and labor costs involved in each of the vessel manufacturing steps. A thorough cost study is performed to understand the SCCV cost as a function of the key vessel design parameters, including hydrogen pressure, vessel dimensions, and load-carrying ratio. The major conclusions include: The SCCV technology can meet the technical/cost targets set forth by DOE s FCT Program for FY2015 and FY2020 for all three pressure levels (i.e., 160, 430 and 860 bar) relevant to the hydrogen production and delivery infrastructure. Further vessel cost reduction can benefit from the development of advanced vessel fabrication technologies such as the highly automated friction stir welding (FSW). The ORNL-patented multi-layer, multi-pass FSW can not only reduce the amount of labor needed for assembling and welding the layered steel vessel, but also make it possible to use even higher strength steels for further cost reductions and improvement of vessel structural integrity. It is noted the cost analysis results demonstrate the significant cost advantage attainable by the SCCV technology for different pressure levels when compared to the industry-standard pressure vessel technology. The real-world performance data of SCCV under actual operating conditions is imperative for this new technology to be adopted by the hydrogen industry for stationary storage of CGH2. Therefore, the key technology development effort in FY13 and subsequent years will be focused on the fabrication and testing of SCCV mock-ups. The static loading and fatigue data will be generated in rigorous testing of these mock-ups. Successful tests are crucial to enabling the near-term impact of the developed storage technology on the CGH2 storage market, a critical component of the hydrogen production and delivery infrastructure. In particular, the SCCV has high potential for widespread deployment in hydrogen fueling stations.

Feng, Zhili [ORNL; Zhang, Wei [ORNL; Wang, Jy-An John [ORNL; Ren, Fei [ORNL

2012-09-01T23:59:59.000Z

108

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect (OSTI)

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

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

109

On-Site Oxy-Lance Size Reduction of South Texas Project Reactor Vessel Heads - 12324  

SciTech Connect (OSTI)

On-Site Oxy-Lance size reduction of mildly radioactive large components has been accomplished at other operating plants. On-Site Oxy-Lance size reduction of more radioactive components like Reactor Vessel Heads had previously been limited to decommissioning projects. Building on past decommissioning and site experience, subcontractors for South Texas Project Nuclear Operating Company (STPNOC) developed an innovative integrated system to control smoke, radioactive contamination, worker dose, and worker safety. STP's innovative, easy to use CEDM containment that provided oxy lance access, smoke control, and spatter/contamination control was the key to successful segmentation for cost-effective and ALARA packaging and transport for disposal. Relative to CEDM milling, STP oxy-lance segmentation saved approximately 40 person- REM accrued during 9,000 hours logged into the radiological controlled area (RCA) during more than 3,800 separate entries. Furthermore there were no personnel contamination events or respiratory uptakes of radioactive material during the course of the entire project. (authors)

Posivak, Edward [WMG, inc. (United States); Keeney, Gilbert; Wheeler, Dean [Shaw Group (United States)

2012-07-01T23:59:59.000Z

110

Reactor Core Assembly - HFIR Technical Parameters | ORNL Neutron Sciences  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Home › Facilities › HFIR › Reactor Core Assembly Home › Facilities › HFIR › Reactor Core Assembly Reactor Core Assembly The reactor core assembly is contained in an 8-ft (2.44-m)-diameter pressure vessel located in a pool of water. The top of the pressure vessel is 17 ft (5.18 m) below the pool surface, and the reactor horizontal mid-plane is 27.5 ft (8.38 m) below the pool surface. The control plate drive mechanisms are located in a subpile room beneath the pressure vessel. These features provide the necessary shielding for working above the reactor core and greatly facilitate access to the pressure vessel, core, and reflector regions. In-core irradiation and experiment locations (cross section at horizontal midplane) Reactor core assembly Reactor core assembly: (1) in-core irradiation and experiment locations,

111

Pressurized water reactor fuel assembly subchannel void fraction measurement  

SciTech Connect (OSTI)

The void fraction measurement experiment of pressurized water reactor (PWR) fuel assemblies has been conducted since 1987 under the sponsorship of the Ministry of International Trade and Industry as a Japanese national project. Two types of test sections are used in this experiment. One is a 5 x 5 array rod bundle geometry, and the other is a single-channel geometry simulating one of the subchannels in the rod bundle. Wide gamma-ray beam scanners and narrow gamma-ray beam computed tomography scanners are used to measure the subchannel void fractions under various steady-state and transient conditions. The experimental data are expected to be used to develop a void fraction prediction model relevant to PWR fuel assemblies and also to verify or improve the subchannel analysis method. The first series of experiments was conducted in 1992, and a preliminary evaluation of the data has been performed. The preliminary results of these experiments are described.

Akiyama, Yoshiei [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan). Nuclear Fuel and Core Engineering Dept.; Hori, Keiichi [Mitsubishi Heavy Industries, Ltd., Hyougo (Japan); Miyazaki, Keiji [Osaka Univ. (Japan). Faculty of Engineering; Mishima, Kaichiro [Kyoto Univ., Osaka (Japan). Research Reactor Inst.; Sugiyama, Shigekazu [Nuclear Power Engineering Corp., Tokyo (Japan). Nuclear Fuel Dept.

1995-12-01T23:59:59.000Z

112

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect (OSTI)

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

113

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

SciTech Connect (OSTI)

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.

Lund, A.L.

1997-11-01T23:59:59.000Z

114

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS  

SciTech Connect (OSTI)

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume 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 does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2010-05-24T23:59:59.000Z

115

OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR  

E-Print Network [OSTI]

OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR K.J. BACHMANN of computer simulations as an optimal design tool which lessens the costs in time and effort in experimental vapor deposition (HPOMCVD) reactor for use in thin film crystal growth. The advantages of such a reactor

116

Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 |  

Broader source: Energy.gov (indexed) [DOE]

Initial Modeling of a Pressurized Water Reactor Completed Using Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 January 29, 2013 - 12:06pm Addthis Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration RELAP-7 is a nuclear reactor system safety analysis code. Development started in October 2011, and during the past quarter the initial capabilities of RELAP-7 were demonstrated by simulating a steady-state single-phase pressurized water reactor (PWR) with two parallel loops and multiple reactor core flow channels (Fig. 1). The PWR configuration matched that of the Three Mile Island 1 LWR, which is a benchmark problem from the

117

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS  

SciTech Connect (OSTI)

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. (2) Minimize the fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. Fill rates that are less than 2 inches/min will reduce the chance of significant hydrogen build-up. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates however, are low for the pH 8 and pH 10.4 grout, i.e., less than 0.32 ft{sup 3}/min. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations. It is recommended that this grout not be utilized for this task. If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2009-10-29T23:59:59.000Z

118

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

SciTech Connect (OSTI)

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

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

2008-05-12T23:59:59.000Z

119

Design and development of a special purpose SAFT system for nondestructive evaluation of nuclear reactor vessels and piping components  

SciTech Connect (OSTI)

This report describes the design details of a special purpose system for real-time nondestructive evaluation of reactor vessels and piping components. The system consists of several components and the report presents the results of the research aimed at the design of each component and recommendations based on the results. One major component of the NDE system, namely the real-time SAFT processor, was designed with sufficient details to enable the fabrications of a prototype by GARD Inc. under a subcontract from The University of Michigan and the report includes their results and conclusions.

Ganapathy, S.; Schmult, B.; Wu, W.S.; Dennehy, T.G.; Moayeri, N.; Kelly, P.

1985-08-01T23:59:59.000Z

120

D0 Silicon Upgrade: Gas Helium Storage Tank Pressure Vessel Engineering Note  

SciTech Connect (OSTI)

This is to certify that Beaird Industries, Inc. has done a white metal blast per SSPC-SP5 as required per specifications on the vessel internal. Following the blast, a black light inspection was performed by Beaird Quality Control personnel to assure that all debris, grease, etc. was removed and interior was clean prior to closing vessel for helium test.

Rucinski, Russ; /Fermilab

1996-11-11T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test  

SciTech Connect (OSTI)

An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication related lack-of-fusion defects, an artificially induced fatigue crack and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach; The welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

Bhuyan, G.S. [Powertech Labs. Inc., Surrey, British Columbia (Canada); Sperling, E.J. [Amoco Corp., Naperville, IL (United States); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Research and Development Corp., Farmers Branch, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

1996-12-01T23:59:59.000Z

122

Prediction of failure behavior of a welded pressure vessel containing flaws during a hydrogen-charged burst test  

SciTech Connect (OSTI)

An industry-government collaborative program was carried out with an aim to promoting the acceptance of fracture mechanics-based fitness-for-service assessment methodology for a service-damaged pressure vessel. A collaborative round robin exercise was carried out to predict the fracture behavior of a vessel containing hydrogen damage, fabrication-related lack-of-fusion defects, an artificially induced fatigue crack, and a localized thinned area. The fracture assessment procedures used include the US ASME Material Property Council`s PREFIS Program based on the British Standard (BS) Published Document (PD) 6493, ASME Section XI and The Central Electricity Generating Board (CEGB) R6 approach, The Welding Institute (TWI) CRACKWISE program (based on BS PD6493 Level 2 approach), a variant of the R6 approach, J-tearing instability approaches, various J-estimation schemes, LEFM approach, and simplified stress analysis. Assessments were compared with the results obtained from a hydrogen-charged burst test of the vessel. Predictions, based on the J-tearing approach, compared well with the actual burst test results. Actual burst pressure was about five times the operating pressure.

Bhuyan, G.S. [Powertech Labs Inc., Surrey, British Columbia (Canada); Sperling, E.J. [BP-Amoco, Calgary, Alberta (Canada); Shen, G. [CANMET, Ottawa, Ontario (Canada). Metals Technology Labs.; Yin, H. [Mobil Technology Co., Dallas, TX (United States); Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States)

1999-08-01T23:59:59.000Z

123

Ion transport membrane module and vessel system  

DOE Patents [OSTI]

An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel. The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); Van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

2012-02-14T23:59:59.000Z

124

Ion transport membrane module and vessel system  

DOE Patents [OSTI]

An ion transport membrane system comprising (a) a pressure vessel having an interior, an exterior, an inlet, and an outlet; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region, wherein any inlet and any outlet of the pressure vessel are in flow communication with exterior regions of the membrane modules; and (c) one or more gas manifolds in flow communication with interior regions of the membrane modules and with the exterior of the pressure vessel.The ion transport membrane system may be utilized in a gas separation device to recover oxygen from an oxygen-containing gas or as an oxidation reactor to oxidize compounds in a feed gas stream by oxygen permeated through the mixed metal oxide ceramic material of the membrane modules.

Stein, VanEric Edward (Allentown, PA); Carolan, Michael Francis (Allentown, PA); Chen, Christopher M. (Allentown, PA); Armstrong, Phillip Andrew (Orefield, PA); Wahle, Harold W. (North Canton, OH); Ohrn, Theodore R. (Alliance, OH); Kneidel, Kurt E. (Alliance, OH); Rackers, Keith Gerard (Louisville, OH); Blake, James Erik (Uniontown, OH); Nataraj, Shankar (Allentown, PA); van Doorn, Rene Hendrik Elias (Obersulm-Willsbach, DE); Wilson, Merrill Anderson (West Jordan, UT)

2008-02-26T23:59:59.000Z

125

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

DOE Patents [OSTI]

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

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

1994-05-03T23:59:59.000Z

126

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

DOE Patents [OSTI]

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

McDermott, Daniel J. (Export, PA); Schrader, Kenneth J. (Penn Hills, PA); Schulz, Terry L. (Murrysville Boro, PA)

1994-01-01T23:59:59.000Z

127

Methane-to-hydrogen conversion in a reversible flow filtration combustion reactor at a high pressure  

Science Journals Connector (OSTI)

The noncatalytic process of partial oxidation of methane to syngas in a reversible flow filtration combustion reactor at high pressures has been considered. ... conversion process — the maximum temperature in the...

Yu. M. Dmitrenko; P. A. Klyovan

2013-09-01T23:59:59.000Z

128

An inverted pressurized water reactor design with twisted-tape swirl promoters  

E-Print Network [OSTI]

An Inverted Fuel Pressurized Water Reactor (IPWR) concept was previously investigated and developed by Paolo Ferroni at MIT with the effort to improve the power density and capacity of current PWRs by modifying the core ...

Nguyen, Nghia T. (Nghia Tat)

2014-01-01T23:59:59.000Z

129

A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors  

SciTech Connect (OSTI)

A model is developed to simulate the oxidation of zircaloy fuel rod cladding exposed to pressurized water reactor operating conditions. The model is used to predict the oxidation rate for both ex- and in-reactor conditions in terms of the weight gain and oxide thickness. Comparisons of the model predictions with experimental data show very good agreement.

Amarshad, A.I.A. [Institute of Atomic Energy Research, Riyadh (Saudi Arabia); Klein, A.C. [Oregon State Univ., Corvallis, OR (United States)

1992-12-31T23:59:59.000Z

130

System pressure effect on the nuclear reactor limiting criterion. Revision 1  

SciTech Connect (OSTI)

The acceptable operating limits of a nuclear reactor are set to prevent fuel cladding damage. Critical Heat Flux (CHF) is the limiting criterion for the high pressure systems such as the BWRs (6.9 MPa) and the PWRs (13.8 MPa). However, the Onset of Flow Instability (OFI) is the limiting criterion of the low pressure system such as the existing Savannah River Site (SRS) production reactors (0.2 MPa). The physical basis of this difference is presented. 3 refs.

Chen, Kuo-Fu

1990-12-31T23:59:59.000Z

131

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

SciTech Connect (OSTI)

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

Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G. [Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushakti Nagar, Mumbai, PIN-400094 (India)

2012-07-01T23:59:59.000Z

132

Ceramic membrane reactor with two reactant gases at different pressures  

DOE Patents [OSTI]

The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

Balachandran, Uthamalingam (Hinsdale, IL); Mieville, Rodney L. (Glen Ellyn, IL)

2001-01-01T23:59:59.000Z

133

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle  

SciTech Connect (OSTI)

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A., E-mail: sedov@dhtp.kial.ru; Subbotin, S. A.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15T23:59:59.000Z

134

High-performance simulations for atmospheric pressure plasma reactor  

Science Journals Connector (OSTI)

Plasma-assisted processing and deposition of materials is an important component of modern industrial applications, with plasma reactors sharing 30% to 40% of manufacturing steps in microelectronics production. Development of new flexible electronics ...

Svyatoslav Chugunov / Iskander Akhatov

2012-01-01T23:59:59.000Z

135

General features of direct-cycle, supercritical-pressure, light-water-cooled reactors  

SciTech Connect (OSTI)

The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR.

Oka, Y.; Koshizuka, S. [Univ. of Tokyo (Japan). Nuclear Engineering Research Lab.

1996-07-01T23:59:59.000Z

136

Characterization of Gas?Liquid Flows in Stirred Vessels Using Pressure and Torque Fluctuations  

Science Journals Connector (OSTI)

Gas?liquid flows in a stirred vessel exhibit different flow regimes and demonstrate complex interaction of transport processes with varying spatio-temporal scales. The knowledge of key space and time scales of fluid dynamics is important for designing and ...

A. R. Khopkar; S. S. Panaskar; A. B. Pandit; V. V. Ranade

2005-03-18T23:59:59.000Z

137

Fossil fuel furnace reactor  

DOE Patents [OSTI]

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

138

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was 50 hours of gasification on a petroleum coke from the Hunt Oil Refinery and an additional 73 hours of operation on a high-ash coal from India. Data from these tests indicate that while acceptable fuel gas heating value was achieved with these fuels, the transport gasifier performs better on the lower-rank feedstocks because of their higher char reactivity. Comparable carbon conversions have been achieved at similar oxygen/coal ratios for both air-blown and oxygen-blown operation for each fuel; however, carbon conversion was lower for the less reactive feedstocks. While separation of fines from the feed coals is not needed with this technology, some testing has suggested that feedstocks with higher levels of fines have resulted in reduced carbon conversion, presumably due to the inability of the finer carbon particles to be captured by the cyclones. These data show that these low-rank feedstocks provided similar fuel gas heating values; however, even among the high-reactivity low-rank coals, the carbon conversion did appear to be lower for the fuels (brown coal in particular) that contained a significant amount of fines. The fuel gas under oxygen-blown operation has been higher in hydrogen and carbon dioxide concentration since the higher steam injection rate promotes the water-gas shift reaction to produce more CO{sub 2} and H{sub 2} at the expense of the CO and water vapor. However, the high water and CO{sub 2} partial pressures have also significantly reduced the reaction of (Abstract truncated)

Michael L. Swanson

2005-08-30T23:59:59.000Z

139

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

SciTech Connect (OSTI)

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

Shaver, Mark W.; Lanning, Donald D.

2010-02-01T23:59:59.000Z

140

HFIR vessel probabilistic fracture mechanics analysis  

SciTech Connect (OSTI)

The life of the High Flux Isotope Reactor (HFIR) pressure vessel is limited by a radiation induced reduction in the material`s fracture toughness. Hydrostatic proof testing and probabilistic fracture mechanics analyses are being used to meet the intent of the ASME Code, while extending the life of the vessel well beyond its original design value. The most recent probabilistic evaluation is more precise and accounts for the effects of gamma as well as neutron radiation embrittlement. This analysis confirms the earlier estimates of a permissible vessel lifetime of at least 50 EFPY (100 MW).

Cheverton, R.D. [Delta-21 Resources, Inc., Oak Ridge, TN (United States); Dickson, T.L. [Oak Ridge National Lab., TN (United States)

1997-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Acoustic emission monitoring of HFIR vessel during hydrostatic testing  

SciTech Connect (OSTI)

This report discusses the results and conclusions reached from applying acoustic emission monitoring to surveillance of the High Flux Isotope Reactor vessel during pressure testing. The objective of the monitoring was to detect crack growth and/or fluid leakage should it occur during the pressure test. The report addresses the approach, acoustic emission instrumentation, installation, calibration, and test results.

Friesel, M.A.; Dawson, J.F.

1992-08-01T23:59:59.000Z

142

The core pressure drop characteristics in a CANDU 6 reactor  

Science Journals Connector (OSTI)

The core pressure drop characteristics in a CANDU 6 have been examined to reveal the mechanism inducing the difference in the core pressure drop among four passages. The general characteristics for the inlet header temperature and core pressure drop are deduced from the measured data of CANDU 6 NPPs. The passages, which are connected to the purification system, are shown to have a larger core pressure drop and lower inlet header temperature compared with other passages in a loop. The temperature difference among four inlet headers has been analytically obtained by considering the effect of the purification system and verified by the measured data of CANDU 6 NPPs. The relationship between the inlet header temperature and core pressure drop has been secured from the magnetite transport mechanism in a CANDU 6. The analytical computations for a CANDU 6 NPP have revealed that the core pressure drop difference among four passages is largely dependent on the single phase friction factor rather than the mass flow rate in a passage. The calculated single phase friction factors are in accord with the magnetite deposition characteristics derived from the difference in the inlet header temperature.

Jun Ho Bae; Jong Yeob Jung

2013-01-01T23:59:59.000Z

143

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect (OSTI)

The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

Michael Swanson; Daniel Laudal

2008-03-31T23:59:59.000Z

144

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts  

SciTech Connect (OSTI)

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs ({approx}$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Duffey, R.B.; Pioro, I.L. [Atomic Energy of Canada, Ltd. (Canada); Gabaraev, B.A.; Kuznetsov, Yu. N. [Research and Development Institute of Power Engineering, ul.M. Krasnoselskaya, 2/8 Moscow, Moscow 107140 (Russian Federation)

2006-07-01T23:59:59.000Z

145

Optimization of Pressurized Oxy4Combustion with Flameless Reactor  

SciTech Connect (OSTI)

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

Malavasi, Massimo; Landegger, Gregory

2014-06-30T23:59:59.000Z

146

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

SciTech Connect (OSTI)

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

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

2015-01-01T23:59:59.000Z

147

An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid  

SciTech Connect (OSTI)

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C. [Ulsan National Inst. of Science and Technology UNIST, 100 Banyeon-ri, Eonyang-eup, Ulju-gun, Ulasn Metropolitan City 689-798 (Korea, Republic of)

2012-07-01T23:59:59.000Z

148

Research and Development Roadmaps for Nondestructive Evaluation of Cables, Concrete, Reactor Pressure Vessels, and Piping Fatigue  

SciTech Connect (OSTI)

The purpose of the Materials Aging and Degradation Pathway is to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on systems, structures, and components is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e., service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enabled by improved methods and techniques for detection, monitoring, and prediction of systems, structures, and components degradation.

Clayton, Dwight A.; Bakhtiari, Sasan; Smith, Cyrus M.; Simmons, Kevin L.; Ramuhalli, Pradeep; Coble, Jamie B.; Brenchley, David L.; Meyer, Ryan M.

2013-04-16T23:59:59.000Z

149

Comparison of actinide production in traveling wave and pressurized water reactors  

SciTech Connect (OSTI)

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

Osborne, A.G.; Smith, T.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas at Austin, Austin, TX (United States)

2013-07-01T23:59:59.000Z

150

Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010  

SciTech Connect (OSTI)

Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

Fort, III, William C.; Kallman, Richard A.; Maes, Miguel; Skolnik, Edward G.; Weiner, Steven C.

2010-12-22T23:59:59.000Z

151

Light Water Reactor Sustainability Technical Documents | Department of  

Broader source: Energy.gov (indexed) [DOE]

Initiatives » Nuclear Reactor Technologies » Light Water Reactor Initiatives » Nuclear Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents September 30, 2011 Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement

152

Development of Improved Composite Pressure Vessels for Hydrogen Storage - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

0 0 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Norman Newhouse (Primary Contact), Jon Knudsen, John Makinson Lincoln Composites, Inc. 5117 NW 40 th Street Lincoln, NE 68524 Phone: (402) 470-5035 Email: nnewhouse@lincolncomposites.com DOE Managers HQ: Ned Stetson Phone: (202) 586-9995 Email: Ned.Stetson@ee.doe.gov GO: Jesse Adams Phone: (720) 356-1421 Email: Jesse.Adams@go.doe.gov Contract Number: DE-FC36-09GO19004 Project Start Date: February 1, 2009 Project End Date: June 30, 2014 Fiscal Year (FY) 2012 Objectives Improve the performance characteristics, including * weight, volumetric efficiency, and cost, of composite pressure vessels used to contain hydrogen in adsorbants. Evaluate design, materials, or manufacturing process *

153

High-pressure reaction and emissions characteristics of catalytic reactors for gas turbine combustors  

Science Journals Connector (OSTI)

The reaction and emissions characteristics of catalytic reactors comprising noble metal catalysts were investigated using homogeneous mixtures of natural gas and vitiated air at pressures up to 2.9 MPa. The mixture temperatures at inlet ranged from 500 to 700°C and the fuel-air ratio was increased till the exit gas temperature reached about 1200°C. Values of combustion efficiency greater than 99.5% and nitrogen oxides emissions for all catalytic reactors tested were less than 0.2 g NO2/kg fuel (2 ppm (15% 02) ) for all reactors at reactor exit gas temperatures higher than about 1100°C. Combustion efficiency decreased with increasing pressure in the heterogeneous-reaction controlled region, though a pressure increase favored homogeneous, gas phase reactions. Appreciable reactivity deterioration by aging for 1000 h at 1000°C was observed at lower mixture temperatures. A two-stage combustor comprising a conventional flame combustion stage and a catalytic stage was fabricated and its NO,x emissions and performance were evaluated at conditions typical of stationary gas turbine combustor operations. About 80% reduction in NO,x emissions levels compared with flame combustion was attained at 1 \\{MPa\\} pressure and 1180°C exit gas temperature, together with complete hydrocarbon combustion.

S. Hayashi; H. Yamada; K. Shimodaira

1995-01-01T23:59:59.000Z

154

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

SciTech Connect (OSTI)

Tests were conducted to study the behavior of Zircaloy fuel cladding under conditions that approximate those found in an operating pressurized-water power reactor. Radial surface displacement values as functions of time, average diametral-circumferential strain as a function of time, and isochronal deformation surfaces are presented. Tests were conducted at 343 degree C with external pressures from 13.3 to 18.7 MPa. Contrary to similar tests conducted out-of-reactor, the in-reactor specimens did not deform uniformly, that is, by diametral contraction and smooth ovalization. Rather, the deformation surfaces were nonuniform with hills and valleys being formed at irregular intervals. This implies that conventional concepts of creep rate and simplified modeling procedures will not work for predicting cladding behavior. 10 refs.

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

1982-01-01T23:59:59.000Z

155

Critical Facility for lattice physics experiments for the Advanced Heavy Water Reactor and the 500 MWe pressurized heavy water reactors  

Science Journals Connector (OSTI)

Bhabha Atomic Research Centre (BARC), Mumbai, is embarking on a broad based program for thorium utilization in power production to achieve all-round capability in the entire thorium cycle. As a step in this direction, a low power Critical Facility is under construction at BARC. The facility will greatly contribute to the understanding and validation of the calculational models and nuclear data used in the design of thorium based Advanced Heavy Water Reactor. The facility is also designed to cater to the experimental requirements of future lattice studies related to 500 MWe pressurized heavy water reactors. This paper covers the basic design features, safety aspects and the planned experimental program of the new facility.

V.K. Raina; R. Srivenkatesan; D.C. Khatri; D.K. Lahiri

2006-01-01T23:59:59.000Z

156

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

SciTech Connect (OSTI)

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

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

157

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - III: Spent DUPIC Fuel Disposal Cost  

SciTech Connect (OSTI)

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Ko, Won Il; Choi, Hangbok; Roh, Gyuhong; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

158

Vertical Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Vertical Pretreatment Reactor System Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment at diff erent temperatures * Biomass is heated by steam injection to temperatures of 120°C to 210°C in the pressurized mixing tube * Preheated, premixed biomass is retained for specified residence time in vertical holding vessel; material continuously moves by gravity from top to bottom of reactor in plug-fl ow fashion * Residence time is adjusted by changing amount of material held in vertical vessel relative to continuous fl ow of material entering and exiting vessel * Optional additional reactor vessel allows for secondary pretreatment at lower temperatures-120°C to 180°C-with potential to add other chemical catalysts * First vessel can operate at residence

159

TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis  

SciTech Connect (OSTI)

The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

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

1986-07-01T23:59:59.000Z

160

Single crystal flow reactor for studying reactivities on metal oxide model catalysts at atmospheric pressure to bridge the pressure gap to the adsorption properties determined under UHV conditions  

Science Journals Connector (OSTI)

A flow reactor for the investigation of heterogeneous catalytic reactions on single crystalline metal oxide model catalysts has been designed. It is located in a high pressure cell attached to an UHV analysis cha...

C. Kuhrs; M. Swoboda; W. Weiss

2001-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Physics characteristics of a large, passive, pressure tube light water reactor with voided calandria  

SciTech Connect (OSTI)

A light water cooled and moderated pressure tube reactor concept has been developed that can survive loss-of-coolant accidents (LOCAs) without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tube. The reactor employs a solid SiC-coated graphite fuel matrix in the pressure tubes and a calandria tank containing a low-pressure gas, surrounded by a graphite reflector. This normally voided calandria is connected to a light water heat sink. The cover gas displaces light water from the calandria during normal operation, while during LOCAs it allows passive calandria flooding. It is shown that such a system, with high void fraction in the core region, exhibits a high degree of neutron thermalization and a large prompt neutron lifetime, similar to D{sub 2}O moderated cores, although light water is used as both coolant and moderator. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile and inherent stability against xenon spatial oscillations. The heterogeneous arrangement of the fuel and moderator ensures a negative void coefficient under all circumstances. Flooding of the calandria space with light water results in redundant reactor shutdown. Use of particle fuel allows attainment of high burnups.

Hejzlar, P.; Driscoll, M.J.; Todreas, N.E. [Massachusetts Inst. of Technology, Cambridge, MA (United States). Dept. of Nuclear Engineering

1995-11-01T23:59:59.000Z

162

Knowledge base expert system control of spatial xenon oscillations in pressurized water reactors  

Science Journals Connector (OSTI)

Current xenon oscillation control methods used in pressurized water reactors are knowledge intensive, and heuristic in nature. An expert system is developed to implement the heuristic constant axial offset control procedure to aid reactor operators in increasing the plant reliability by reducing the human error component of the failure probability. An expert system is written in the production system language, OPS5, with a forward chaining algorithm. It samples the reactor core with a certain time interval, evaluates the core status to determine the necessary corrective actions in terms of reactivity insertion, and selects the control parameter to realize this reactivity insertion. The amount of control action is determined using a knowledge base which consists of the differential rod worth curves, and the boron reactivity worth of a given reactor. The controller has been tested using a one-dimesional core model for verification of the rules and the code. It has been shown that, having the reactor dependent parameters in its knowledge base, the controller is able to follow a typical load demand for a daily cycle of a reactor, and is able to keep the axial offset within a target band.

Serhat Alten; Richard A Danofsky

1993-01-01T23:59:59.000Z

163

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

SciTech Connect (OSTI)

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

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

1981-09-15T23:59:59.000Z

164

The Heating of Electrons in Magnetic Traps of Low-Pressure Electron-Cyclotron-Resonance Microwave-Frequency Reactors  

Science Journals Connector (OSTI)

Methods are analyzed of maintaining plasma in low-pressure electron-cyclotron-resonance reactors in which the ionization is ... The key part played by zones of electron-cyclotron heating of electrons by the elect...

A. B. Petrin

165

Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors  

E-Print Network [OSTI]

Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, ...

Hannink, Ryan Christopher

2007-01-01T23:59:59.000Z

166

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

SciTech Connect (OSTI)

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

Todosow M.; Todosow M.; Raitses, G. (BNL) Galperin, A. (Ben Gurion University)

2009-07-12T23:59:59.000Z

167

Acoustic emission monitoring of HFIR vessel during hydrostatic testing. Final report  

SciTech Connect (OSTI)

This report discusses the results and conclusions reached from applying acoustic emission monitoring to surveillance of the High Flux Isotope Reactor vessel during pressure testing. The objective of the monitoring was to detect crack growth and/or fluid leakage should it occur during the pressure test. The report addresses the approach, acoustic emission instrumentation, installation, calibration, and test results.

Friesel, M.A.; Dawson, J.F.

1992-08-01T23:59:59.000Z

168

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

SciTech Connect (OSTI)

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

Hobson, D.O. (Oak Ridge National Lab., TN); Thoms, K.R.; Dodd, C.V.; van der Kaa, Th.

1982-01-01T23:59:59.000Z

169

Nuclear reactor  

DOE Patents [OSTI]

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

170

Attrition reactor system  

DOE Patents [OSTI]

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

171

Effects of Hyperbaric Pressure on a Deep-Sea Archaebacterium in Stainless Steel and Glass-Lined Vessels  

Science Journals Connector (OSTI)

...syringe and the pressure generator provides ballast to...turn of the pressure generator. As with gas sam...these experiments, water was used as the pressurizing...microorganisms from water and sediment samples...subsequent incuba- tion at atmospheric pressure. In our experiments...

Chad M. Nelson; Michael R. Schuppenhauer; Douglas S. Clark

1991-12-01T23:59:59.000Z

172

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

SciTech Connect (OSTI)

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

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

173

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

SciTech Connect (OSTI)

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

Goldberg, A.; Streit, R.D.

1981-05-01T23:59:59.000Z

174

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

SciTech Connect (OSTI)

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

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

2008-04-16T23:59:59.000Z

175

E-Print Network 3.0 - aged reactor pressure Sample Search Results  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

effort Summary: plan to build a 5 billion fusion reactor, called the International Thermonuclear Experimental Reactor... could be achieved in 35 years. "By the time our young...

176

Multivent effects in a large scale boiling water reactor pressure suppression system  

SciTech Connect (OSTI)

The steam-driven GKSS pressure suppression test facility, which contains 3 full scale vent pipes, has been used for 5 years to investigate the postulated loss-of-coolant accident in a Mark II and Type 69 boiling water reactor. Using the results from several of these tests, wetwell boundary load data (peak pressures and spectral power) during the chugging stage, have been evaluated for sparse pool response (one and two vents in the three vent pool) and for full pool response (one, two, or three vent operation in pools of constant wetwell pool area per vent). The sparse pool results indicate the pool-system, chug event boundary loads are strongly dependent on wetwell pool area per vent, with the load increasing with decreasing area. The full pool results show a substantial increase in the pool-system, chug event boundary loads upon a change from single cell to double cell operation; only minor change occurs in going from double to triple cell operation.

McCauley, E.W.; Aust, E.; Schwan, H.

1984-07-06T23:59:59.000Z

177

Application of LBB to high energy pipings of a pressurized water reactor in Korea  

Science Journals Connector (OSTI)

The application of the leak before break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved in Korea for several high energy systems that can meet rigorous acceptance criteria. The LBB application in Korea is based on the US Nuclear Regulatory Commision (USNRC) requirements. The purpose of the LBB evaluation is to eliminate the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to eliminate pipe whip restraints and jet impingement barriers. There were several issues on the application of LBB to the primary coolant loop and the pressurizer surge line. Of concern were the material properties for the carbon steel for the primary coolant loop, estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry, and the leakage crack size which barely meets the required margin of 2 for the surge line, etc. Some additional work was required by the safety authority to maintain the global safety of the plant at a sufficient level. This paper describes the regulatory application of LBB in Korea, and the issues encountered during the regulatory review.

Jeong-Bae Lee; Young Hwan Choi

1999-01-01T23:59:59.000Z

178

Optimized Adaptive Fuzzy Controller of the Water Level of a Pressurized Water Reactor Steam Generator  

Science Journals Connector (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

M. Marseguerra; E. Zio; F. Cadini

179

Device for automating in vitro characterization of lymphatic vessel function  

E-Print Network [OSTI]

. 22 Figure 14 exhibits the relationship between peak pressure and Emax for the three vessels. As peak pressure increases, Emax increases monotonically for vessel 3. Vessel 3 exhibits a linear relationship (R2=0.92). The trend is less apparent.... 22 Figure 14 exhibits the relationship between peak pressure and Emax for the three vessels. As peak pressure increases, Emax increases monotonically for vessel 3. Vessel 3 exhibits a linear relationship (R2=0.92). The trend is less apparent...

Rajagopalan, Shruti

2005-02-17T23:59:59.000Z

180

Gas-cooled nuclear reactor  

DOE Patents [OSTI]

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

1985-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

System and method for determining coolant level and flow velocity in a nuclear reactor  

DOE Patents [OSTI]

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10T23:59:59.000Z

182

Effects of light water reactor coolant environment on the fatigue lives of  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Effects of light water reactor coolant environment on the fatigue lives of Effects of light water reactor coolant environment on the fatigue lives of reactor materials July 8, 2013 A metal component can become progressively degraded, and its structural integrity can be adversely impacted when it is subjected to repeated fluctuating loads, or fatigue loading. Fatigue loadings on nuclear reactor pressure vessel components can occur because of changes in pressure and temperature caused by transients during operation, such as reactor startup or shutdown and turbine trip events. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code recognizes fatigue as a possible cause of failure of reactor materials and provides rules for designing nuclear power plant components to avoid fatigue failures. For various materials, the ASME Code defines the

183

NAPTH: Neutronics analysis code system for the fusion–fission hybrid reactor with pressure tube type blanket  

Science Journals Connector (OSTI)

The fusion–fission hybrid reactor is considered as a potential path to the early application of fusion energy. A new concept with pressure tube type blanket has recently been proposed for a feasible hybrid reactor. In this paper, a code system for the neutronics analysis of the pressure tube type hybrid reactor is developed based on the two-step calculation scheme: the few-group homogeneous constant calculation and the full blanket calculation. The few-group homogeneous constants are calculated using the lattice code DRAGON4. The blanket transport calculation is performed by the multigroup Monte Carlo code. A link procedure for fitting the cross sections is developed between these two steps. An additional procedure is developed to calculate the burnup, power distribution, energy multiplication factor, tritium breeding ratio and neutron multiplication factor. From some numerical results, it is found that the code system NAPTH is reliable and exhibits good calculation efficiency. It can be used for the conceptual design of the pressure tube type hybrid reactor with precise geometry.

Tiejun Zu; Hongchun Wu; Youqi Zheng; Liangzhi Cao; Chao Yang

2013-01-01T23:59:59.000Z

184

The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures  

E-Print Network [OSTI]

shaft. With the impeller in place and rotating, gas was drawn into the top port and ejected at the impeller mount. The reactor pressure was monitored via the transducer port. The transducer was a Viatran Pressure Transducer, model 103. The liquid...THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial...

Hess, Richard Kenneth

2012-06-07T23:59:59.000Z

185

Fracture toughness evaluation for N Reactor pressure tubes of Zircaloy-2 using the electric-potential method  

SciTech Connect (OSTI)

Zircaloy is commonly used for the cladding or pressure tubes in commercial reactors because of its strength, corrosion resistance, and low absorption of thermal neutrons. Fracture toughness test techniques using small samples fabricated from archival materials from N Reactor pressure tubes of Zircaloy-2 were developed to study the factors affecting tube fracture toughness. Compact tension specimen thickness was limited by the wall thickness (7 mm) of the tubes. Specimens (5 mm thick) were prepared for fracture toughness testing and results were analyzed using the J-integral approach. To reduce the high cost of irradiated specimen testing and to more easily precrack specimens remotely, single-specimen potential drop techniques were employed to evaluate the fracture toughness of Zircaloy-2. The initiation fracture toughness was determined from J-R curves, which were constructed by plotting values of J as a function of crack extension computed from the electric-potential calibration curve. The J-R curves obtained from the calibration curve equation fit the blunting line region. The curves also fit heat-tint data well. The effects of neutron fluence and hydrogen content on the fracture toughness of N Reactor pressure tubes were evaluated. Neutron irradiation substantially degraded fracture toughness. Increasing fluence decreased the fracture toughness of the alloy. Hydrogen also decreased fracture toughness, but this effect was insignificant for the pressure tubes tested. 8 refs., 13 figs., 3 tabs.

Huang, F.H.

1991-11-01T23:59:59.000Z

186

An Advanced Computational Scheme for the Optimization of 2D Radial Reflectors in Pressurized Water Reactors  

E-Print Network [OSTI]

This paper presents a computational scheme for the determination of equivalent 2D multi-group heterogeneous reflectors in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de donn\\'{e}es et Aide \\`{a} l'Optimisation (ADAO)" of the SALOME platform developed at \\'{E}lectricit\\'{e} De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first validation of the computational scheme is made using the OPTEX reflector model developed at \\'{E}cole Polytechnique de Montr\\'{e}al (EPM). As a result, we obtain 2D multi-group, spatially heterogeneous 2D reflectors, using both diffusion or $\\text{SP}_{\\text{N}}$ operators. We observe important improvements of the power discrepancies distribution over the cor...

Clerc, Thomas; Leroyer, Hadrien; Argaud, Jean-Philippe; Bouriquet, Bertrand; Ponçot, Agélique

2014-01-01T23:59:59.000Z

187

SAFT and TOFD—A Comparative Study of Two Defect Sizing Techniques on a Reactor Pressure Vessel Mock-up  

Science Journals Connector (OSTI)

Defect sizing is required for a quantitative assessment of the quality and reliability of safety relevant components and materials using ultrasonic non-destructive testing. The SAFT (Synthetic Aperture Focussing

Jens Prager; Jessica Kitze; Cécile Acheroy…

2013-03-01T23:59:59.000Z

188

Failure Analysis, Permeation, and Toughness of Glass Fiber Composite Pressure Vessels for Inexpensive Delivery of Cold Hydrogen - DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

2 2 DOE Hydrogen and Fuel Cells Program FY 2012 Annual Progress Report Andrew Weisberg (Primary Contact), Salvador Aceves Lawrence Livermore National Laboratory (LLNL) P.O. Box 808, L-792 Livermore, CA 94551 Phone: (925) 422-0864 Email: saceves@llnl.gov DOE Manager HQ: Erika Sutherland Phone: (202) 586-3152 Email: Erika.Sutherland@ee.doe.gov Subcontractor: Spencer Composites Corporation (SCC), Sacramento, CA Project Start Date: October, 2004 Project End Date: October, 2012 Fiscal Year (FY) 2012 Objectives Optimize hydrogen delivery by tube trailer * Develop materials and manufacturing for low- * temperature hydrogen delivery Quantify performance and economics of delivery * pressure vessels Technical Barriers This project addresses the following technical barriers

189

A combustion/deposition entrained reactor for high?temperature/pressure studies of coal and coal minerals  

Science Journals Connector (OSTI)

The combustion of coal and coal?derived fuels in heat engines poses significant technical challenges in terms of establishing high combustion rates and efficiencies controlling emissions and minimizing the impact of fuel contaminants on engine components. An entrained reactor has been designed and constructed to study coal particle combustion the tendency of coal ash to form deposits on heat engine components and the effects of fuel additives on residual ash composition. The reactor is designed for high temperature/pressure conditions similar to those of a coal?fired gas turbine. Optical access ports and advanced instrumentation allow the i n s i t u measurement of gas and particle temperatures and vapor phase alkali concentrations. The reactor has been used to study the deposition potential of several coals as a function of process conditions and to determine the effects of selected additives on the deposition rate.

Rodney J. Anderson; Ronald G. Logan; Charles T. Meyer; Richard A. Dennis

1990-01-01T23:59:59.000Z

190

CO2 Gasification Rates of Petroleum Coke in a Pressurized Flat-Flame Burner Entrained-Flow Reactor  

Science Journals Connector (OSTI)

Two petcoke samples were gasified by CO2 at total pressures of 10 and 15 atm in a high-pressure flat-flame burner reactor at conditions where the bulk phase consisted of either 40 or 90 mol % CO2 with gas temperatures up to 1909 K. Particle diameters of 45–75 ?m were used in the experiments. ... The mass release data caused by CO2 gasification of the petcoke chars were fit to a global first-order model, and the optimal kinetic parameters are reported. ... The CO2 char gasification rates of both petcokes were shown to be higher than Illinois #6 coal when reacted at conditions of high temperature and pressure, even though most reactivity comparisons between petcoke and coal at lower temperature, pressure, and heating rates typically result in coal being more reactive. ...

Aaron D. Lewis; Emmett G. Fletcher; Thomas H. Fletcher

2014-06-05T23:59:59.000Z

191

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

SciTech Connect (OSTI)

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

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

192

An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors  

SciTech Connect (OSTI)

An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

Gentry, Cole A [ORNL] [ORNL; Godfrey, Andrew T [ORNL] [ORNL; Terrani, Kurt A [ORNL] [ORNL; Gehin, Jess C [ORNL] [ORNL; Powers, Jeffrey J [ORNL] [ORNL; Maldonado, G Ivan [ORNL] [ORNL

2014-01-01T23:59:59.000Z

193

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

Science Journals Connector (OSTI)

Abstract A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 ?m and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (?0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235U enrichment and the additional UO2 pellet volume enabled by using thinner cladding.

Nathan Michael George; Kurt Terrani; Jeff Powers; Andrew Worrall; Ivan Maldonado

2015-01-01T23:59:59.000Z

194

Thermal hydraulic performance analysis of a small integral pressurized water reactor core  

E-Print Network [OSTI]

A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three ...

Blair, Stuart R. (Stuart Ryan), 1972-

2003-01-01T23:59:59.000Z

195

Subchannel Thermal-Hydraulic Experimental Program (STEP). Volume 1. Mixing in a pressurized water reactor (PWR) rod bundle. Final report  

SciTech Connect (OSTI)

This volume describes an experiment that was performed to determine the mixing characteristics of a pressurized water reactor (PWR) rod bundle. The objective of this project was to improve the subchannel computer code models of the reactor core. The experimental technique was isokinetic subchannel withdrawal of the entire flow from two sample subchannels. Once withdrawn, the sample fluid was condensed and its enthalpy was measured by regenerative heat exchange calorimetry. The test bundle was a 4 x 6 electrically heated array with a 50% power upset. The COBRA IIIC code was used to model the experiment and to determine the value of the thermal mixing coefficient, ..beta.., that was necessary to predict the measured results. Both single- and two-phase data were obtained over a range of PWR operating conditions. The results indicate that both single- and two-phase mixing is small. The COBRA model predicts the enthalpy data using a turbulent mixing coefficient, ..beta.. approx. = 0.002.

Barber, A.R.; Zielke, L.A.

1980-08-01T23:59:59.000Z

196

Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents  

SciTech Connect (OSTI)

During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ?5% of maximum value by the final moment of the stage of linear increase in the temperature.

Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M. [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)] [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)

2013-09-11T23:59:59.000Z

197

ALARA Analysis for Shippingport Pressurized Water Reactor Core 2 Fuel Storage in the Canister Storage Building (CSB)  

E-Print Network [OSTI]

The addition of Shippingport Pressurized Water Reactor (PWR) Core 2 Blanket Fuel Assembly storage in the Canister Storage Building (CSB) will increase the total cumulative CSB personnel exposure from receipt and handling activities. The loaded Shippingport Spent Fuel Canisters (SSFCs) used for the Shippingport fuel have a higher external dose rate. Assuming an MCO handling rate of 170 per year (K East and K West concurrent operation), 24-hr CSB operation, and nominal SSFC loading, all work crew personnel will have a cumulative annual exposure of less than the 1,000 mrem limit.

Lewis, M E

2000-01-01T23:59:59.000Z

198

Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building  

SciTech Connect (OSTI)

This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

Lata

1996-09-26T23:59:59.000Z

199

Analysis of a natural circulation cooldown transients in a Westinghouse Pressurized Water Reactor using TRAC-PF1/MOD1 and TRAC-PF1/MOD2  

E-Print Network [OSTI]

Circulation Cooldown Transient in a Westinghouse Pressurized Water Reactor Using TRAC-PF1/MOD1 and TRAC-PF1/MOD2. (December 1988) Evelyn Marie Breiner, B. S. , Texas AgtM University Chair of Advisory Committee; Dr. B. Nassersharif To perform transient.... 22). The four-loop model differs from the two-loop 35 TABLE 5 Component Actuation Timing Component Action Transient Time (s) 4-Loo Mod 1 Transient Time (s) 2-Loo M 1 4" break occurs CVCS initiation Low pressurizer pressure trip Reactor trip...

Breiner, Evelyn Marie

1988-01-01T23:59:59.000Z

200

Atmospheric Pressure Weakly Ionized Plasma Reactor Based on the Corona Discharge .  

E-Print Network [OSTI]

??Atmospheric pressure weakly ionized plasma (APWIP) is being used to treat or process goods and materials because it only activates the surface without modification of… (more)

[No author

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report  

SciTech Connect (OSTI)

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

Lancaster, D.

1992-10-26T23:59:59.000Z

202

Sliding Mode Control for Pressurized-Water Nuclear Reactors in load following operations with bounded xenon oscillations  

Science Journals Connector (OSTI)

Abstract One of the important operations in nuclear power plants is load-following in which imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation considered to be a constraint for the load-following operation. In this paper, sliding mode control (SMC) which is a robust nonlinear controller is designed to control the Pressurized-Water Nuclear Reactor (PWR) power for the load-following operation problem that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to maintain xenon oscillations to be bounded. The constant AO is a robust state constraint for load-following problem. The reactor core is simulated based on the two-point nuclear reactor model and one delayed neutron group. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Results show that the proposed controller for the load-following operation is sufficiently effective so that the xenon oscillations are kept bounded in the considered region.

G.R. Ansarifar; S. Saadatzi

2015-01-01T23:59:59.000Z

203

Low-temperature rupture behavior of Zircaloy clad pressurized water reactor spent fuel rods under dry storage conditions  

SciTech Connect (OSTI)

Creep rupture studies on five well-characterized Zircaloy clad pressurized water reactor spent fuel rods, which were pressurized to a hoop stress of approximately 145 MPa, were conducted for up to 2101 h at 323/sup 0/C. The conditions were chosen for limited annealing of in-reactor irradiation-hardening. No cladding breaches occurred, although significant hydride agglomeration and reorientation took place in rods that cooled under stress. Observations are interpreted in terms of a conservatively modified Larson-Miller curve to provide a lower bound on permissible maximum dry-storage temperatures, assuming creep rupture as the life-limiting mechanism. If hydride reorientation can be ruled out during dry storage, 305/sup 0/C is a conservative lower bound, based on the creep rupture mechanism, for the maximum storage temperature of rods with irradiation hardened cladding to ensure a 100-year cladding lifetime in an inert atmosphere. An oxidizing atmosphere reduces the lower bound on the maximum permissible storage temperature by approx. 5/sup 0/C. While high-temperature tests based on creep rupture as the limiting mechanism indicate that storage at temperatures between 400/sup 0/C and 440/sup 0/C may be feasible for rods which are annealed, tests to study rod performance in the 305/sup 0/ to 400/sup 0/C temperature range have not been conducted. 37 references, 10 figures, 7 tables.

Einziger, R.E.; Kohli, R.

1983-01-01T23:59:59.000Z

204

Low-temperature rupture behavior of Zircaloy-clad pressurized water reactor spent fuel rods under dry storage conditions  

SciTech Connect (OSTI)

Creep rupture studies on five well-characterized Zircaloy-clad pressurized water reactor spent fuel rods, which were pressurized to a hoop stress of about145 MPa, were conducted for up to 2101 h at 323/sup 0/C. The conditions were chosen for limited annealing of in-reactor irradiation hardening. No cladding breaches occurred, although significant hydride agglomeration and reorientation took place in rods that cooled under stress. Observations are interpreted in terms of a conservatively modified Larson-Miller curve to provide a lower bound on permissible maximum dry-storage temperatures, assuming creep rupture as the life-limiting mechanism. If hydride reorientation can be ruled out during dry storage, 305/sup 0/C is a conservative lower bound, based on the creep-rupture mechanism, for the maximum storage temperature of rods with irradiation-hardened cladding to ensure a 100-yr cladding lifetime in an inert atmosphere. An oxidizing atmosphere reduced the lower bound on the maximum permissible storage temperature by about5/sup 0/C. While this lower bound is based on whole-rod data, other types of data on spent fuel behavior in dry storage might support a higher limit. This isothermal temperature limit does not take credit for the decreasing rod temperature during dry storage. High-temperature tests based on creep rupture as the limiting mechanism indicate that storage at temperatures between 400 and 440/sup 0/C may be feasible for rods that are annealed.

Einsiger, R.E.; Kohli, R.

1984-10-01T23:59:59.000Z

205

Computed phase equilibria for burnable neutron absorbing materials for advanced pressurized heavy water reactors  

Science Journals Connector (OSTI)

Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU®[CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.

E.C. Corcoran; B.J. Lewis; W.T. Thompson; J. Hood; F. Akbari; Z. He; P. Reid

2009-01-01T23:59:59.000Z

206

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

SciTech Connect (OSTI)

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

Miglin, M.T.; Monter, J.V.; Wade, C.S. [Babcock & Wilcox Co., Alliance, OH (United States); Nelson, J.L. [Electric Power Research Institute, Palo Alto, CA (United States)

1992-12-31T23:59:59.000Z

207

Development of an internally cooled annular fuel bundle for pressurized heavy water reactors  

SciTech Connect (OSTI)

A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01T23:59:59.000Z

208

Fuel handling apparatus for a nuclear reactor  

DOE Patents [OSTI]

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

209

Combustion synthesis continuous flow reactor  

DOE Patents [OSTI]

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

1998-01-01T23:59:59.000Z

210

High-Pressure Tube Trailers and Tanks  

Broader source: Energy.gov (indexed) [DOE]

bending stress: continuous fiber vessels and vessels made of replicants Conformable tanks require internal stiffeners (ribs) to efficiently support the pressure and minimize...

211

Heat dissipating nuclear reactor with metal liner  

DOE Patents [OSTI]

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

212

Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview  

SciTech Connect (OSTI)

Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work.

Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

1987-09-01T23:59:59.000Z

213

Final report for confinement vessel analysis. Task 2, Safety vessel impact analyses  

SciTech Connect (OSTI)

This report describes two sets of finite element analyses performed under Task 2 of the Confinement Vessel Analysis Program. In each set of analyses, a charge is assumed to have detonated inside the confinement vessel, causing the confinement vessel to fail in either of two ways; locally around the weld line of a nozzle, or catastrophically into two hemispheres. High pressure gases from the internal detonation pressurize the inside of the safety vessel and accelerate the fractured nozzle or hemisphere into the safety vessel. The first set of analyses examines the structural integrity of the safety vessel when impacted by the fractured nozzle. The objective of these calculations is to determine if the high strength bolt heads attached to the nozzle penetrate or fracture the lower strength safety vessel, thus allowing gaseous detonation products to escape to the atmosphere. The two dimensional analyses predict partial penetration of the safety vessel beneath the tip of the penetrator. The analyses also predict maximum principal strains in the safety vessel which exceed the measured ultimate strain of steel. The second set of analyses examines the containment capability of the safety vessel closure when impacted by half a confinement vessel (hemisphere). The predicted response is the formation of a 0.6-inch gap, caused by relative sliding and separation between the two halves of the safety vessel. Additional analyses with closure designs that prevent the gap formation are recommended.

Murray, Y.D. [APTEK, Inc., Colorado Springs, CO (United States)

1994-01-26T23:59:59.000Z

214

Evaluation of cracking in feedwater piping adjacent to the steam generators in Nine Pressurized Water Reactor Plants  

SciTech Connect (OSTI)

Cracking in ASTM A106-B and A106-C feedwater piping was detected near the inlet to the steam generators in a number of pressurized water reactor plants. We received sections with cracks from nine of the plants with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Variations were observed in piping surface irregularities, corrosion-product, pit, and crack morphology, surface elmental and crystal structure analyses, and steel microstructures and mechanical properties. However, with but two exceptions, namely, arrest bands and major surface irregularities, we were unable to relate the extent of cracking to any of these factors. Tensile and fracture toughness (J/sub Ic/ and tearing modulus) properties were measured over a range of temperatures and strain rates. No unusual properties or microstructures were observed that could be related to the cracking problem. All crack surfaces contained thick oxide deposits and showed evidence of cyclic events in the form of arrest bands. Transmission electron microscopy revealed fatigue striations on replicas of cleaned crack surfaces from one plant and possibly from three others. Calculations based on the observed striation spacings gave a value of ..delta..sigma = 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses. Although surface irregularities and corrosion pits were sources for crack initiation and corrosion may have contributed to crack propagation, it is proposed that the overriding factor in the cracking problem is the presence of unforeseen cyclic loads.

Goldberg, A.; Streit, R.D.; Scott, R.G.

1980-06-25T23:59:59.000Z

215

Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning  

SciTech Connect (OSTI)

This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

Shah, V.N.; Ware, A.G.; Porter, A.M.

1997-03-01T23:59:59.000Z

216

Experimental and Modeling Studies of the Methane Steam Reforming Reaction at High Pressure in a Ceramic Membrane Reactor.  

E-Print Network [OSTI]

??This dissertation describes the preparation of a novel inorganic membrane for hydrogen permeation and its application in a membrane reactor for the study of the… (more)

Hacarlioglu, Pelin

2007-01-01T23:59:59.000Z

217

Radiation embrittlement of PWR vessel supports  

SciTech Connect (OSTI)

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100/degree/C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs.

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01T23:59:59.000Z

218

Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

International Hydrogen International Hydrogen Fuel and Pressure Vessel Forum to someone by E-mail Share Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on Facebook Tweet about Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on Twitter Bookmark Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on Google Bookmark Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on Delicious Rank Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on Digg Find More places to share Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure Vessel Forum on AddThis.com... Publications Program Publications Technical Publications

219

Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies  

SciTech Connect (OSTI)

A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application.

Ham, Y S; Maldonado, G I; Burdo, J; He, T

2006-10-10T23:59:59.000Z

220

Estimation of the xenon concentration and delayed neutrons precursors densities in the pressurized-water nuclear reactors (PWR) with sliding mode observer considering xenon oscillations  

Science Journals Connector (OSTI)

Abstract One of the important operations in nuclear power plants is load-following in which the imbalance in axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load-following operation. On the other hands, precursors produce delayed neutrons which are important with respect to reactor period and control, but xenon concentration and precursors densities cannot be measured directly. In this paper, the non-linear sliding mode observer which has the robust characteristics facing the parameters uncertainties and disturbances is proposed based on the two point nuclear reactor model equations with three groups of the delayed neutrons to estimate the xenon concentration and delayed neutrons precursor densities of the pressurized-water nuclear reactor (PWR) using reactor power measurements. The stability analysis is provided by means of the Lyapunov approach, thus the system is guaranteed to be stable over a wide range. The employed method is easy to implement. This estimation is done taking into account the effects of reactivity feedback due to temperature and xenon concentration. Simulation results clearly show that the sliding mode observer follows the actual system variables accurately and is satisfactory in the presence of the parameter uncertainties and disturbances.

M.H. Esteki; G.R. Ansarifar; M. Arghand

2015-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions  

SciTech Connect (OSTI)

A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup ?4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

Morfin, Franck; Piccolo, Laurent [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)] [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

2013-09-15T23:59:59.000Z

222

USING LIGA BASED MICROFABRICATION TO IMPROVE OVERALL HEAT TRANSFER EFFICIENCY OF PRESSURIZED WATER REACTOR: I. Effects of Different Micro Pattern on Overall Heat Transfer.  

SciTech Connect (OSTI)

The Pressurized Water Reactors (PWRs in Figure 1) were originally developed for naval propulsion purposes, and then adapted to land-based applications. It has three parts: the reactor coolant system, the steam generator and the condenser. The Steam generator (a yellow area in Figure 1) is a shell and tube heat exchanger with high-pressure primary water passing through the tube side and lower pressure secondary feed water as well as steam passing through the shell side. Therefore, a key issue in increasing the efficiency of heat exchanger is to improve the design of steam generator, which is directly translated into economic benefits. The past research works show that the presence of a pin-fin array in a channel enhances the heat transfer significantly. Hence, using microfabrication techniques, such as LIGA, micro-molding or electroplating, some special microstructures can be fabricated around the tubes in the heat exchanger to increase the heat-exchanging efficiency and reduce the overall size of the heat-exchanger for the given heat transfer rates. In this paper, micro-pin fins of different densities made of SU-8 photoresist are fabricated and studied to evaluate overall heat transfer efficiency. The results show that there is an optimized micro pin-fin configuration that has the best overall heat transfer effects.

Zhang, M.; Ibekwe, S.; Li, G.; Pang, S.S.; and Lian, K.

2006-07-01T23:59:59.000Z

223

R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels | Department...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Hydrogen Fuel and Pressure Vessel Forum Bonfire Tests of High Pressure Hydrogen Storage Tanks Status and Progress in Research, Development and Demonstration of Hydrogen-Compressed...

224

E-Print Network 3.0 - abnormal blood vessels Sample Search Results  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

of Summary: ) the blood vessels, which also helps to lower blood pressure. Commonly used brand names in the United States... to treat high blood pressure, heart disease and...

225

TR-EDB: Test Reactor Embrittlement Data Base, Version 1  

SciTech Connect (OSTI)

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States)

1994-01-01T23:59:59.000Z

226

Ex-Core CFD Analysis Results for the Prometheus Gas Reactor  

SciTech Connect (OSTI)

This paper presents the initial nozzle-to-nozzle (N2N) reactor vessel model scoping studies using computational fluid dynamics (CFD) analysis methods. The N2N model has been solved under a variety of different boundary conditions. This paper presents some of the basic hydraulic results from the N2N CFD analysis effort. It also demonstrates how designers were going to apply the analysis results to modify a number of the design features. The initial goals for developing a preliminary CFD N2N model were to establish baseline expectations for pressure drops and flow fields around the reactor core. Analysis results indicated that the averaged reactor vessel pressure drop for all analyzed cases was 46.9 kPa ({approx}6.8 psid). In addition, mass flow distributions to the three core fuel channel regions exhibited a nearly inverted profile to those specified for the in-core thermal/hydraulic design. During subsequent design iterations, the goal would have been to modify or add design features that would have minimized reactor vessel pressure drop and improved flow distribution to the inlet of the core.

Lorentz, Donald G. [Space Engineering, Bechtel Bettis, Inc. West Mifflin, PA 15122 (United States)

2007-01-30T23:59:59.000Z

227

New fast-reactor approach. [LMFBR  

SciTech Connect (OSTI)

The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel.

Folkrod, J.R.; Kann, W.J.; Klocksieben, R.H.

1983-01-01T23:59:59.000Z

228

Coal gasification vessel  

DOE Patents [OSTI]

A vessel system (10) comprises an outer shell (14) of carbon fibers held in a binder, a coolant circulation mechanism (16) and control mechanism (42) and an inner shell (46) comprised of a refractory material and is of light weight and capable of withstanding the extreme temperature and pressure environment of, for example, a coal gasification process. The control mechanism (42) can be computer controlled and can be used to monitor and modulate the coolant which is provided through the circulation mechanism (16) for cooling and protecting the carbon fiber and outer shell (14). The control mechanism (42) is also used to locate any isolated hot spots which may occur through the local disintegration of the inner refractory shell (46).

Loo, Billy W. (Oakland, CA)

1982-01-01T23:59:59.000Z

229

EDS V25 containment vessel explosive qualification test report.  

SciTech Connect (OSTI)

The V25 containment vessel was procured by the Project Manager, Non-Stockpile Chemical Materiel (PMNSCM) as a replacement vessel for use on the P2 Explosive Destruction Systems. It is the first EDS vessel to be fabricated under Code Case 2564 of the ASME Boiler and Pressure Vessel Code, which provides rules for the design of impulsively loaded vessels. The explosive rating for the vessel based on the Code Case is nine (9) pounds TNT-equivalent for up to 637 detonations. This limit is an increase from the 4.8 pounds TNT-equivalency rating for previous vessels. This report describes the explosive qualification tests that were performed in the vessel as part of the process for qualifying the vessel for explosive use. The tests consisted of a 11.25 pound TNT equivalent bare charge detonation followed by a 9 pound TNT equivalent detonation.

Rudolphi, John Joseph

2012-04-01T23:59:59.000Z

230

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5  

SciTech Connect (OSTI)

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k{sub eff} of 1. 0040{+-}0.0005.

Bowman, S.M. [Oak Ridge National Lab., TN (United States); Suto, T. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)]|[Oak Ridge National Lab., TN (United States)

1996-10-01T23:59:59.000Z

231

Characterization of Reactor Ash Deposits from Pilot-Scale Pressurized Entrained-Flow Gasification of Woody Biomass  

Science Journals Connector (OSTI)

(21, 22) Second, crude syngases can carry fly ash particles and gaseous inorganics that cause detrimental effects downstream, such as corrosion of heat recovery systems, contamination of syngas cleaning sorbents, and poisoning of biofuel production catalysts. ... Milled fuel is fed into the reactor via a single top-fired centrally positioned burner. ... The near-wall temperatures attained in the PEFG reactor were around 200–300 °C higher than the bed temperatures of most FB gasifiers/combustors.(4, 47-49) Considering also the temporal temperature elevations of particles (e.g., vicinal burning particles and passage through the O2-enriched flame regions), this would mean that both the average and temporal peak particle temperatures are significantly higher during PEFG. ...

Charlie Ma; Fredrik Weiland; Henry Hedman; Dan Boström; Rainer Backman; Marcus Öhman

2013-11-04T23:59:59.000Z

232

Ion transport membrane module and vessel system with directed internal gas flow  

DOE Patents [OSTI]

An ion transport membrane system comprising (a) a pressure vessel having an interior, an inlet adapted to introduce gas into the interior of the vessel, an outlet adapted to withdraw gas from the interior of the vessel, and an axis; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region; and (c) one or more gas flow control partitions disposed in the interior of the pressure vessel and adapted to change a direction of gas flow within the vessel.

Holmes, Michael Jerome (Thompson, ND); Ohrn, Theodore R. (Alliance, OH); Chen, Christopher Ming-Poh (Allentown, PA)

2010-02-09T23:59:59.000Z

233

Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for  

Broader source: Energy.gov (indexed) [DOE]

Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping Light water reactor sustainability (LWRS) nondestructive evaluation (NDE) Workshops were held at Oak Ridge National Laboratory (ORNL) during July 30th to August 2nd, 2012. This activity was conducted to help develop the content of the NDE R&D roadmap for the materials aging and degradation (MAaD) pathway of the LWRS program. The workshops focused on identifying NDE R&D needs in four areas: cables, concrete, reactor pressure vessel, and piping. A selected group of subject matter experts (SMEs) from DOE national

234

VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0  

SciTech Connect (OSTI)

Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

Duo, J. I. [Radiation Engineering and Analysis, Westinghouse Electric Company LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2011-07-01T23:59:59.000Z

235

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

SciTech Connect (OSTI)

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

Palmer, M. E.; Fielder, R. S.; Davis, M. A. [Luna Innovations, Incorporated, 2851 Commerce St., Blacksburg, VA 24060 (United States)

2006-07-01T23:59:59.000Z

236

Investigation of downward facing critical heat flux with water-based nanofluids for In-Vessel Retention applications  

E-Print Network [OSTI]

In-Vessel Retention ("IVR") is a severe accident management strategy that is power limiting to the Westinghouse AP1000 due to critical heat flux ("CHF") at the outer surface of the reactor vessel. Increasing the CHF level ...

DeWitt, Gregory L

2011-01-01T23:59:59.000Z

237

Gamma dose from activation of internal shields in IRIS reactor  

Science Journals Connector (OSTI)

......located inside the reactor vessel, which requires...the sake of better reliability and safety, it...located inside the reactor vessel, which requires...the sake of better reliability and safety, it...Equipment Failure Analysis methods European...Carlo Method Nuclear Reactors Radiation Dosage......

Stefano Agosteo; Antonio Cammi; Luisella Garlati; Carlo Lombardi; Enrico Padovani

2005-12-20T23:59:59.000Z

238

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

SciTech Connect (OSTI)

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

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

1988-07-01T23:59:59.000Z

239

E-Print Network 3.0 - arch vessel transposition Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

the effect of applying local vacuum pressure on the temperatures of the epidermis and small vessels during... of skin and blood vessels with different diameters (10-60 mm) at...

240

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report  

SciTech Connect (OSTI)

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

Mac Donald, Philip Elsworth

2002-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Investigations on optimization of accident management measures following a station blackout accident in a VVER-1000 pressurized water reactor  

SciTech Connect (OSTI)

The reactor safety issues are of primary importance for preserving the health of the population and ensuring no release of radioactivity and fission products into the environment. A part of the nuclear research focuses on improvement of the safety of existing nuclear power plants. Studies, research and efforts are a continuing process at improving the safety and reliability of existing and newly developed nuclear power plants at prevention of a core melt accident. Station blackout (loss of AC power supply) is one of the dominant accidents taken into consideration at performing accident analysis. In case of multiple failures of safety systems it leads to a severe accident. To prevent an accident to turn into a severe one or to mitigate the consequences, accident management measures must be performed. The present paper outlines possibilities for application and optimization of accident management measures following a station blackout accident. Assessed is the behaviour of the nuclear power plant during a station blackout accident without accident management measures and with application of primary/secondary side oriented accident management measures. Discussed are the possibilities for operators ' intervention and the influence of the performed accident management measures on the course of the accident. Special attention has been paid to the effectiveness of the passive feeding and physical phenomena having an influence on the system behaviour. The performed simulations show that the effectiveness of the secondary side feeding procedure can be limited due to an early evaporation or flashing effects in the feed water system. The analyzed cases show that the effectiveness of the accident management measures strongly depends on the initiation criteria applied for depressurization of the reactor coolant system. (authors)

Tusheva, P.; Schaefer, F.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden (Germany)

2012-07-01T23:59:59.000Z

242

Cryogenic Pressure Vessels: Progress and Plans  

Broader source: Energy.gov [DOE]

Presented at the R&D Strategies for Compressed, Cryo-Compressed and Cryo-Sorbent Hydrogen Storage Technologies Workshops on February 14 and 15, 2011.

243

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof  

DOE Patents [OSTI]

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

1991-01-01T23:59:59.000Z

244

Bonfire Tests of High Pressure Hydrogen Storage Tanks  

Broader source: Energy.gov (indexed) [DOE]

Bonfire Tests of High Pressure Hydrogen Storage Tanks International Hydrogen Fuel and Pressure Vessel Forum 2010Beijing, P.R. China September 27, 2010 Bonfire Tests of High...

245

Advanced Nuclear Reactors | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

246

Catalytic reactor  

DOE Patents [OSTI]

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

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

247

Review of the proposed materials of construction for the SBWR and AP600 advanced reactors  

SciTech Connect (OSTI)

Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F. [Argonne National Lab., IL (United States)

1994-06-01T23:59:59.000Z

248

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents [OSTI]

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1990-01-01T23:59:59.000Z

249

Impacts of reducing shipboard NOx? and SOx? emissions on vessel performance  

E-Print Network [OSTI]

The international maritime community has been experiencing tremendous pressures from environmental organizations to reduce the emissions footprint of their vessels. In the last decade, air emissions, including nitrogen ...

Caputo, Ronald J., Jr. (Ronald Joseph)

2010-01-01T23:59:59.000Z

250

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

SciTech Connect (OSTI)

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

Zhang, D.; Rahnema, F. [Georgia Inst. of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

2012-07-01T23:59:59.000Z

251

Tow Vessel | Open Energy Information  

Open Energy Info (EERE)

Vessel Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleTowVessel&oldid596390" Category: Hydrodynamic Testing Facility Type...

252

Pressure testing of torispherical heads  

SciTech Connect (OSTI)

Two vessels fabricated from SA516-70 steel with 6% knuckle radius torispherical heads were tested under internal pressure to failure. The D/t ratios of Vessel 1 and Vessel 2 were 238 and 185 respectively. The calculated maximum allowable working pressures of Vessel 1 and 2 heads using the ASME Section 8, Div. 1 rules and measured dimensions were 85 and 110 psi, respectively. Vessel 1 failed at a nozzle weld in the cylindrical shell at 700 psi pressure. Neither buckling nor any other objectionable deformation of the head was observed at a theoretical double-elastic-slope collapse pressure of 241 and a calculated buckling pressure of 270 psi. Buckles were observed developing slowly after 600 psi pressure, and a total of 22 buckles were observed after the test, having the maximum amplitude of 0.15 inch. Vessel 2 failed at the edge of the longitudinal weld of the cylindrical shell at 1,080 psi pressure. Neither buckling nor any other objectionable deformation of the head was observed up to the final pressure, which exceeded the theoretical double-elastic-slope collapse and calculated buckling pressures of 274 psi and 342 psi, respectively.

Rana, M.D. [Praxair, Inc., Tonawanda, NY (United States). Research and Development Dept.; Kalnins, A.; Updike, D.P. [Lehigh Univ., Bethlehem, PA (United States)

1995-12-01T23:59:59.000Z

253

Application of Computational Physics: Blood Vessel Constrictions and Medical Infuses  

E-Print Network [OSTI]

Application of computation in many fields are growing fast in last two decades. Increasing on computation performance helps researchers to understand natural phenomena in many fields of science and technology including in life sciences. Computational fluid dynamic is one of numerical methods which is very popular used to describe those phenomena. In this paper we propose moving particle semi-implicit (MPS) and molecular dynamics (MD) to describe different phenomena in blood vessel. The effect of increasing the blood pressure on vessel wall will be calculate using MD methods, while the two fluid blending dynamics will be discussed using MPS. Result from the first phenomenon shows that around 80% of constriction on blood vessel make blood vessel increase and will start to leak on vessel wall, while from the second phenomenon the result shows the visualization of two fluids mixture (drugs and blood) influenced by ratio of drugs debit to blood debit. Keywords: molecular dynamic, blood vessel, fluid dynamic, movin...

Suprijadi,; Subekti, Petrus; Viridi, Sparisoma

2013-01-01T23:59:59.000Z

254

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

SciTech Connect (OSTI)

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (high pressure) and LPP9 (low pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be studied. A probabilistic risk analysis was performed that indicated that the probability of occurrence for these two transients is on the order of 10[sup [minus]5] and 10[sup [minus]7] per reactor year for LCP15 and LPP9, respectively. The fission product behavior analysis included calculations of the gaseous and highly volatile fission product (xenon, krypton, cesium, iodine, and tellurium) inventories in the fuel before accident initiation, release of the fission products from the fuel into the reactor vessel during core melt, the probable chemical forms, and transport of the fission products from the core through the reactor vessel and existing piping to the confinement. In addition to a base-case analysis of fission product behavior, a series of analyses was performed to determine the sensitivity of fission product release to several parameters including steam flow rate, (structural) aluminum oxidation, and initial aerosol size. The base-case analyses indicate that the volatile fission products (excluding the noble gases) will be transported as condensed species on zinc aerosols.

Adams, J.P.; Carboneau, M.L.; Hagrman, D.L. (Idaho National Engineering Lab. EG and G Idaho, Idaho Falls, ID (United States))

1993-07-01T23:59:59.000Z

255

Reconnecting broken blood vessels  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Reconnecting broken blood vessels Reconnecting broken blood vessels Name: Catherine A Kraft Status: N/A Age: N/A Location: N/A Country: N/A Date: N/A Question: While watching the television program "Chicago Hope" the other day, I watched a doctor sew someone's ear back on using an elaborate microscope. I was wondering if a surgeon is required to reconnect all the broken blood vessels, and how you would accomplish this? Thanks for your time! Replies: I'm not a surgeon, but I think the answer to your question is "no." The blood will flow across the wound (out the end of one blood vessel and into the end of another), although not efficiently. I believe they sometimes use leeches sucking on the end of the reconnected part to help induce flow of blood in the right direction through the area. You probably do need to put the ends of the major vessels near each other, so the distribution of blood flow is reasonably like it was before the injury, and so the vessels can eventually reconnect. But probably the microscope is used mostly to be sure the various layers of muscle, connective tissue, and fat are connected together correctly.

256

Liquid metal cooled nuclear reactor plant system  

DOE Patents [OSTI]

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

257

An Acoustically Driven Magnetized Target Fusion Reactor  

Science Journals Connector (OSTI)

We propose a new compression system that offers many advantages. A near spherical vessel ?2 m in diameter is filled with liquid lead-lithium alloy (PbLi). This liquid is under consideration for fusion reactor bla...

Michel Laberge

2008-06-01T23:59:59.000Z

258

Containment system for supercritical water oxidation reactor  

DOE Patents [OSTI]

A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

Chastagner, P.

1994-07-05T23:59:59.000Z

259

Weld monitor and failure detector for nuclear reactor system  

DOE Patents [OSTI]

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

Sutton, Jr., Harry G. (Mt. Lebanon, PA)

1987-01-01T23:59:59.000Z

260

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Surveillance Guide - OSS 19.4 Pressure Safety  

Broader source: Energy.gov (indexed) [DOE]

PRESSURE SAFETY PRESSURE SAFETY 1.0 Objective The objective of this surveillance is to evaluate the contractor's implementation of programs to ensure the integrity of pressure vessels and minimize risks from failure of vessels to the public and to workers. Facility Representatives will examine the installed configuration of pressure vessels, observe pressure testing and review documentation associated with maintenance or repair of pressure vessels. In performing the surveillance, Facility Representatives will examine implementation of applicable DOE requirements and best practices. 2.0 References 2.1 DOE 5480.4, Environmental Protection, Safety and Health Protection Standards 2.2 DOE 5483.1A, Occupational Safety and Health Programs

262

Microsoft Word - power_reactors_briggs.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and as it moves past the fuel, it carries away the heat produced by the

263

Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for  

Broader source: Energy.gov (indexed) [DOE]

(LWRS) Program - R&D Roadmap (LWRS) Program - R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping Light water reactor sustainability (LWRS) nondestructive evaluation (NDE) Workshops were held at Oak Ridge National Laboratory (ORNL) during July 30th to August 2nd, 2012. This activity was conducted to help develop the content of the NDE R&D roadmap for the materials aging and degradation (MAaD) pathway of the LWRS program. The workshops focused on identifying NDE R&D needs in four areas: cables, concrete, reactor pressure vessel, and piping. A selected group of subject matter experts (SMEs) from DOE national laboratories, academia, vendors, EPRI, and NRC were invited to each

264

Assessing the Effects of Radiation Damage on Ni-base Alloys for the Prometheus Space Reactor System  

SciTech Connect (OSTI)

Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron ,exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculations show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the Be0 control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.

T. Angeliu

2006-01-19T23:59:59.000Z

265

Assessing the Effects of Radiation Damage on Ni-base Alloys for the Prometheus Space Reactor System  

SciTech Connect (OSTI)

Ni-base alloys were considered for the Prometheus space reactor pressure vessel with operational parameters of {approx}900 K for 15 years and fluences up to 160 x 10{sup 20} n/cm{sup 2} (E > 0.1 MeV). This paper reviews the effects of irradiation on the behavior of Ni-base alloys and shows that radiation-induced swelling and creep are minor considerations compared to significant embrittlement with neutron exposure. While the mechanism responsible for radiation-induced embrittlement is not fully understood, it is likely a combination of helium embrittlement and solute segregation that can be highly dependent on the alloy composition and exposure conditions. Transmutation calculations show that detrimental helium levels would be expected at the end of life for the inner safety rod vessel (thimble) and possibly the outer pressure vessel, primarily from high energy (E > 1 MeV) n,{alpha} reactions with {sup 58}Ni. Helium from {sup 10}B is significant only for the outer vessel due to the proximity of the outer vessel to the BeO control elements. Recommendations for further assessments of the material behavior and methods to minimize the effects of radiation damage through alloy design are provided.

T Angeliu; J Ward; J Witter

2006-04-04T23:59:59.000Z

266

Usage Codes Vessel name  

E-Print Network [OSTI]

Usage Codes 1 5 2 6 3 7 4 8 Vessel name Int'l radio call sign (IRCS) Generator Other: Max hoisting Sonar Power (Kw) KHz: KHz: VMS Usage Y / N GPS: Internal / external KHz: KHz: Ratio Accuracy (m Incinerator: Burned on board: Net sensors Hull mounted / towed Wired / wireless Y / N Y / N Usage Manufacturer

267

NETL: News Release - Ocean Research Vessel Returns with Undersea 'Treasure'  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

23, 2002 23, 2002 Ocean Research Vessel Returns with Undersea 'Treasure' of Methane Hydrates Largest Amount of Marine Hydrate Core Ever Recovered - The R/V JOIDES Resolution - The R/V JOIDES Resolution VICTORIA, BRITISH COLUMBIA - An internationally funded ocean research vessel has returned to port after a two-month expedition off the Oregon coast, bringing with it the largest amount of marine methane hydrate core samples ever recovered for scientific study. The R/V JOIDES Resolution, the world's largest scientific drillship, docked at Victoria, British Columbia earlier this month and began offloading pressure vessels containing methane hydrates recovered 50 miles offshore of Oregon from an area known as Hydrate Ridge. The pressure vessels, each six feet long and four inches in diameter, will

268

Method for destroying hazardous organics and other combustible materials in a subcritical/supercritical reactor  

DOE Patents [OSTI]

A waste destruction method using a reactor vessel to combust and destroy organic and combustible waste, including the steps of introducing a supply of waste into the reactor vessel, introducing a supply of an oxidant into the reactor vessel to mix with the waste forming a waste and oxidant mixture, introducing a supply of water into the reactor vessel to mix with the waste and oxidant mixture forming a waste, water and oxidant mixture, reciprocatingly compressing the waste, water and oxidant mixture forming a compressed mixture, igniting the compressed mixture forming a exhaust gas, and venting the exhaust gas into the surrounding atmosphere.

Janikowski, Stuart K. (Idaho Falls, ID)

2000-01-01T23:59:59.000Z

269

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production  

SciTech Connect (OSTI)

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01T23:59:59.000Z

270

HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR  

E-Print Network [OSTI]

HEART AND BLOOD VESSELS CARDIOVASCULARCARDIOVASCULAR SYSTEMSYSTEM SYSTEM COMPONENTS · Heart pumps blood though blood vessels where exchanges can take place with the interstitial fluid (between cells) · Heart and blood vessels regulate blood flow according to the needs of the body

Cochran-Stafira, D. Liane

271

Fatigue and environmentally assisted cracking in light water reactors  

SciTech Connect (OSTI)

Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

1992-03-01T23:59:59.000Z

272

Control system for a small fission reactor  

DOE Patents [OSTI]

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

1985-02-08T23:59:59.000Z

273

Nuclear reactor downcomer flow deflector  

DOE Patents [OSTI]

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

2011-02-15T23:59:59.000Z

274

DHCVIM: A direct heating containment vessel interactions module  

SciTech Connect (OSTI)

Models for prediction of direct containment heating phenomena as implemented in the DHCVIM computer module are described. The models were designed to treat thermal, chemical and hydrodynamic processes in the three regions of the Sandia National Laboratory Surtsey DCH test facility: the melt generator, cavity and vessel. The fundamental balance equations, along with constitutive relations are described. A combination of Eulerian treatment for the gas phase and Lagrangian treatment for the droplet phase is used in the modeling. Comparisons of calculations and DCH-1 test results are presented. Reasonable agreement is demonstrated for the vessel pressure rise, melt generator pressure decay and particle size distribution.

Ginsberg, T.; Tutu, N.K.

1987-01-01T23:59:59.000Z

275

A Carbon Dioxide Gas Turbine Direct Cycle with Partial Condensation for Nuclear Reactors  

SciTech Connect (OSTI)

A carbon dioxide gas turbine power generation system with a partial condensation cycle has been proposed for thermal and fast nuclear reactors, in which compression is done partly in the liquid phase and partly in the gas phase. This cycle achieves higher cycle efficiency than a He direct cycle mainly due to reduced compressor work of the liquid phase and of the carbon dioxide real gas effect, especially in the vicinity of the critical point. If this cycle is applied to a thermal reactor, efficiency of this cycle is about 55% at a reactor outlet temperature of 900 deg. C and pressure of 12.5 MPa, which is higher by about 10% than a typical helium direct gas turbine cycle plant (PBMR) at 900 deg. C and 8.4 MPa; this cycle also provides comparable cycle efficiency at the moderate core outlet temperature of 600 deg. C with that of the helium cycle at 900 deg. C. If this cycle is applied to a fast reactor, it is anticipated to be an alternative to liquid metal cooled fast reactors that can provide slightly higher cycle efficiency at the same core outlet temperature; it would eliminate safety problems, simplify the heat transport system and simplify plant maintenance. A passive decay heat removal system is realized by connecting a liquid carbon dioxide storage tank with the reactor vessel and by supplying carbon dioxide gasified from the tank to the core in case of depressurization event. (authors)

Yasuyoshi Kato; Takeshi Nitawaki; Yoshio Yoshizawa [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2002-07-01T23:59:59.000Z

276

High Flux Isotope Reactor system RELAP5 input model  

SciTech Connect (OSTI)

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

Morris, D.G.; Wendel, M.W.

1993-01-01T23:59:59.000Z

277

High-Pressure Tube Trailers and Tanks  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Berry Berry Salvador M. Aceves Lawrence Livermore National Laboratory (925) 422-0864 saceves@LLNL.GOV DOE Delivery Tech Team Presentation Chicago, Illinois February 8, 2005 Inexpensive delivery of compressed hydrogen with ambient temperature or cryogenic compatible vessels * Pressure vessel research at LLNL Conformable (continuous fiber and replicants) Cryo-compressed * Overview of delivery options * The thermodynamics of compressed and cryo-compressed hydrogen storage * Proposed analysis activities * Conclusions Outline We are investigating two techniques for reduced bending stress: continuous fiber vessels and vessels made of replicants Conformable tanks require internal stiffeners (ribs) to efficiently support the pressure and minimize bending stresses Spherical and cylindrical tanks

278

Reactor Safety Research Programs  

SciTech Connect (OSTI)

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.

Edler, S. K.

1981-07-01T23:59:59.000Z

279

The siting of UK nuclear reactors  

Science Journals Connector (OSTI)

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

Malcolm Grimston; William J Nuttall; Geoff Vaughan

2014-01-01T23:59:59.000Z

280

Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR (High Flux Isotope Reactor) Reactor  

SciTech Connect (OSTI)

The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs.

Childs, R.L.; Rhoades, W.A.; Williams, L.R.

1988-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

STRESS WAVE EMISSION AND FRACTURE OF PRESTRESSED CONCRETE REACTOR...  

Office of Scientific and Technical Information (OSTI)

PRESTRESSED CONCRETE REACTOR VESSEL MATERIALS. Re-direct Destination: Temp Data Fields Green, A.T. Temp Data Storage 3: Aerojet-General Corp., Sacramento, Calif. Short URL for...

282

High-Pressure Model Catalyst System | EMSL  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

of one atmosphere or below in a reactor situated just below an ultra-high vacuum (UHV) chamber. In particular, heterogeneous catalytic reactions at realistic pressures over...

283

Analysis of a 4-inch small-break loss-of-coolant accident in a Westinghouse Pressurized Water Reactor using TRAC-PF1/MOD1  

E-Print Network [OSTI]

the transient response of a Westinghouse 4- loop PWR using 17x17 fuel assemblies in a 14-ft. long reactor core to a 4-inch diameter SBLOCA with the computer code TRAC-PF1/MOD1, This is unique in that there are only two Westinghouse PWRs with 14-ft. cores (The... 4-inch SBLOCAs 65 XI. Comparison of RESAR-3S, TRAC and RELAP SBLOCAs . . 70 LIST OF ACRONYMS Acronym Name CCFL CVCS ECCS EPRI FSAR HPI INEL LB LOCA LOCA LPI MSIV NRC PCT PORV PWR RCP RCS RESAR RHR SI SBLOCA Argonne National...

Knippel, Kimberley I.R.

2012-06-07T23:59:59.000Z

284

Acoustic Emission Monitoring of ASME Section III Hydrostatic Test: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect (OSTI)

Through the cooperation of the Tennessee Valley Authority, Pacific Northwest Laboratory has installed instrumentation on Watts Bar Nuclear Power Plant Unit 1 for the purpose of test and evaluation of acoustic emission (AE) monitoring of nuclear reactor pressure vessels and piping for flaw detection. This report describes the acoustic emission monitoring performed during the ASME Section III hydrostatic testing of Watts Bar Nuclear Power Plant Unit 1 and the results obtained. Highlights of the results are: • Spontaneous AE was detected from a nozzle area during final pressurization. • Evaluation of the apparent source of the spontaneous AE using an empirically derived AE/fracture mechanics relationship agreed within a factor of two with an evaluation by ASME Section XI Code procedures. • AE was detected from a fracture specimen which was pressure coupled to the 10-inch accumulator nozzle. This provided reassurance of adequate system sensitivity. • High background noise was observed when all four reactor coolant pumps were operating. Work is continuing at Watts Bar Unit 1 toward AE monitoring hot functional testing and subsequently monitoring during reactor operation.

Hutton,, P. H.; Taylor,, T. T.; Dawson,, J. F.; Pappas,, R. A.; Kurtz,, R. J.

1982-06-01T23:59:59.000Z

285

R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels  

Broader source: Energy.gov (indexed) [DOE]

hydrogen accelerates crack propagation rate of the material and leads to brittle fracture. International Hydrogen Fuel and Pressure Vessel Forum 2010Beijing, P.R. China R&D...

286

Research on hydrogen environment fatigue test system and correlative fatigue test of hydrogen storage vessel  

Science Journals Connector (OSTI)

A 70MPa hydrogen environment fatigue test system has been designed and applied in the manufacture of a hydrogen storage vessel. Key equipment is the 80MPa flat steel ribbon wound high pressure hydrogen storage ve...

Rong Li ? ?; Chuan-xiang Zheng ???…

2014-02-01T23:59:59.000Z

287

Reactor for exothermic reactions  

DOE Patents [OSTI]

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

288

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents [OSTI]

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, P.R.; McLennan, G.A.

1984-08-30T23:59:59.000Z

289

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents [OSTI]

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, Paul R. (Western Springs, IL); McLennan, George A. (Downers Grove, IL)

1985-01-01T23:59:59.000Z

290

Hydrogasification reactor and method of operating same  

SciTech Connect (OSTI)

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10T23:59:59.000Z

291

Spherical torus fusion reactor  

DOE Patents [OSTI]

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

292

CRAD, Pressurized Systems and Cryogens Assessment Plan  

Broader source: Energy.gov [DOE]

Assure personnel health and safety through regularly scheduled inspections and maintenance on pressure vessels and equipment, compressed gases and gas cylinders, vacuum equipment and systems, hydraulics, and cryogenic materials and systems.

293

The construction of the Browns Bay Vessel  

E-Print Network [OSTI]

INVESTIGATIVE TECHNIQUES. 10 19 The Site. National Historic Sites Service Excavation and Raising of the Vessel Vessel on Display. The Vessel in 1985. 19 20 27 28 Method of Recording III THE CONSTRUCTION OF THE VESSEL 31 36 The Keel 36 The Stem... A flat-bottomed boat being built. 17 9 Forelocked eye-bolts from the midship beam of the Browne Bay Vessel 21 10 Broad arrow stamped in an eye-bolt from the Browns Bay Vessel. . . . . . . . . . . . . . . . . . . . . . . . . . . 22 11 Pulley...

Amer, Christopher Francis

2012-06-07T23:59:59.000Z

294

E-Print Network 3.0 - axi-symmetric exit pressures Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

driven by the imposed high pressure at the top and the lower exit pressure. The inner tube provides... 1 Paper submitted to ASME 2008 ASME Pressure Vessels and Piping ... Source:...

295

Multidimensional shielding analysis of the JASPER in-vessel fuel storage experiments  

SciTech Connect (OSTI)

The In-Vessel Fuel Storage (IVFS) experiments analyzed in this report were conducted at the Oak Ridge National Laboratory`s Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present report describes the 2-D and 3-D models, analyses, and calculated results corresponding to a limited subset of those IVFS experiments in which the US LMR program has a particular interest.

Bucholz, J.A.

1993-03-01T23:59:59.000Z

296

Predicting Stenosis in Blood Vessels  

E-Print Network [OSTI]

of plaque Plaque is made up of cholesterol, calcium, and other blood components that stick to the vessel-flow loop is function of degree of stenosis, even at low degrees of stenosis So, stenosis may be detected

Petta, Jason

297

Materials Reliability Program Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111)  

SciTech Connect (OSTI)

Over the last thirty years, stress corrosion cracking in PWR primary water (PWSCC) has been observed in numerous Alloy 600 component items and associated welds, sometimes after relatively long incubation times. Repairs and replacements have generally utilized wrought Alloy 690 material and its compatible weld metals (Alloy 152 and Alloy 52), which have been shown to be very highly resistant to PWSCC in laboratory experiments and have been free from cracking in operating reactors over periods already up to nearly 15 years. It is nevertheless prudent for the PWR industry to attempt to quantify the longevity of these materials with respect to aging degradation by corrosion in order to provide a sound technical basis for the development of future inspection requirements for repaired or replaced component items. This document first reviews numerous laboratory tests, conducted over the last two decades, that were performed with wrought Alloy 690 and Alloy 52 or Alloy 152 weld materials under various test conditions pertinent to corrosion resistance in PWR environments. The main focus of the present review is on PWSCC, but secondary-side conditions are also briefly considered.

H.Xu, S.Fyfitch, P.Scott, M.Foucault, R.Kilian, and M.Winters

2004-03-01T23:59:59.000Z

298

High pressure synthesis gas conversion. Task 3: High pressure profiles  

SciTech Connect (OSTI)

The purpose of this research project was to build and test a high pressure fermentation system for the production of ethanol from synthesis gas. The fermenters, pumps, controls, and analytical system were procured or fabricated and assembled in our laboratory. This system was then used to determine the effects of high pressure on growth and ethanol production by C. 1jungdahlii. The limits of cell concentration and mass transport relationships were found in CSTR and immobilized cell reactors (ICR). The minimum retention times and reactor volumes were found for ethanol production in these reactors.

Not Available

1993-05-01T23:59:59.000Z

299

Torsional ultrasonic technique for reactor vessel liquid level measurement  

SciTech Connect (OSTI)

We have undertaken a detailed study of an ultrasonic waveguide employed as a level, density, and temperature sensor. The purpose of this study was to show how such a device might be used in the nuclear power industry to provide reliable level information with a multifunction sensor, thus overcomming several of the errors that led to the accident at Three Mile Island. Some additional work is needed to answer the questions raised by the current study, most noticably the damping effects of flowing water.

Dress, W.B.

1983-01-01T23:59:59.000Z

300

Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools  

E-Print Network [OSTI]

The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents...

Frisani, Angelo

2011-08-08T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Estimation of the sub-criticality of the sodium-cooled fast reactor Monju using the modified neutron source multiplication method  

SciTech Connect (OSTI)

The Modified Neutron Source Method (MNSM) is applied to the Monju reactor. This static method to estimate sub-criticality has already given good results on commercial Pressurized Water Reactors. The MNSM consists both in the extraction of the fundamental mode seen by a detector to avoid the effect of higher modes near sources, and the correction of flux distortion effects due to control rod movement. Among Monju's particularities that have a big influence on MNSM factors are: the presence of two californium sources and the position of the detector which is located far from the core outside of the reactor vessel. The importance of spontaneous fission and ({alpha}, n) reactions which have increased during the shutdown period of 15 years will also be discussed. The relative position of detectors and sources deeply affect the correction factors in some regions. In order to evaluate the detector count rate, an analytical propagation has been conducted from the reactor vessel. For two subcritical states, an estimation of the reactivity has been made and compared to experimental data obtained in the restart experiments at Monju (2010). (authors)

Truchet, G. [Institut National des Sciences et Techniques Nucleaires, Centre CEA de Saclay, F-91191 Gif-sur-Yvette Cedex (France); Van Rooijen, W. F. G.; Shimazu, Y. [Research Inst. of Nuclear Engineering, Univ. of Fukui, Kanawa-cho, 1-2-4, T 914-0055, Fukui-ken, Tsuruga-shi (Japan); Yamaguchi, K. [Japan Atomic Energy Agency, FBR Plant Engineering Center, 919-1279 Fukui-ken, Tsuruga-shi Shiraki 1 (Japan)

2012-07-01T23:59:59.000Z

302

Nuclear reactor with low-level core coolant intake  

DOE Patents [OSTI]

A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

1993-01-01T23:59:59.000Z

303

Fracture Analysis of Vessels – Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations  

SciTech Connect (OSTI)

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include the NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels – Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.

Williams, P. T. [ORNL; Dickson, T. L. [ORNL; Yin, S. [ORNL

2007-12-01T23:59:59.000Z

304

Former Reactor Facilities Surveillance and Maintenance and  

E-Print Network [OSTI]

and Dark (2010) Majority of remaining radiation contained in reactor vessel (inside biological shield window damage Minor roof leaks to former office areas Stack and Grounds: Stack drain tank Safety of confinement dome Replaced exhaust fan Security system hardware/software upgrade BGRR: Covered vents above 2

Ohta, Shigemi

305

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......condition that the reactor was operated at full...Coolant water 55.0 Reactor vessel 30.0 Containment...at utilisation of nuclear energy other than electricity generation(3). Two PSRD units of 100 MW thermal reactor power are located......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

306

Machine for monitoring nuclear reactor equipment at its storage station  

SciTech Connect (OSTI)

Machine for monitoring internal equipment of a nuclear reactor, when stored outside the reactor vessel in the swimming pool of the reactor, comprising a first longitudinal carriage movable on rails on the bottom of the swimming pool, a transverse beam movable on the first carriage, a second transverse carriage bearing monitoring instruments and movable on the beam, and an arrangement for moving the beam in the predetermined steps on the first carriage, and the second carriage on the beam.

Castrec, Y. M.; Launay, J. P.

1985-02-12T23:59:59.000Z

307

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents [OSTI]

An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Hunsbedt, Anstein (Los Gatos, CA)

1996-01-01T23:59:59.000Z

308

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents [OSTI]

An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

Hunsbedt, A.

1996-03-12T23:59:59.000Z

309

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings  

Broader source: Energy.gov (indexed) [DOE]

experts presented information and data on testing and certification of storage tanks for compressed hydrogen, CNG, and HCNG fuels. 1 Specific objectives of the Forum were...

310

Light water reactor lower head failure analysis  

SciTech Connect (OSTI)

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-10-01T23:59:59.000Z

311

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect (OSTI)

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Neil Todreas; Pavel Hejzlar

2008-06-30T23:59:59.000Z

312

NUCLEAR REACTORS.  

E-Print Network [OSTI]

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

313

Development of a Simulation Model and Safety Evaluation for a Depressurization Accident Without Reactor Scram in an Advanced HTGR  

SciTech Connect (OSTI)

It is important to use analyses to prove outstanding inherent reactor safety during a severe accident in order to convince the public and licensing authority of the safety advantage of the high-temperature gas-cooled reactor (HTGR). In this study, the simulation of a depressurization accident without reactor scram (DAWS) was performed for a future HTGR with 450-MW thermal output, introducing the annular core of pin-in-block-type fuel, which was originally designed in Japan. The DAWS has the possibility of becoming one of the severe accidents postulated in the HTGR. To perform an accurate simulation, a new analytical model for reactor dynamics and indirect decay heat removal from the surface of the reactor pressure vessel (RPV) during the DAWS was developed. The features of the new simulation model are as follows:1. A single-channel model is coupled with a two-dimensional reactor thermal model in the new simulation model. The reactor kinetics with a single-channel model during the DAWS is simulated taking into account heat removal from the reactor calculated in the R-Z reactor thermal model, including the RPV and indirect vessel cooling system. No conventional calculation codes with a single channel have a heat removal model from an RPV or were able to simulate precisely the transient during DAWS.2. A xenon buildup and decay model for the reactivity calculation is made in addition to one point-kinetics approximation to simulate a recriticality and a power oscillation following the initiation of the DAWS.3. A transient simulation can be performed for two kinds of core models of pin-in-block- and multihole-type fuels.The accurate evaluation of xenon density and core temperature is of prime importance in the simulation of the DAWS. From the simulation result with a proper safety margin, it was confirmed that the safety performance of passive decay heat removal with cooling indirectly from the surface of the RPV is outstanding for the DAWS, and a severe-accident-free HTGR can be designed. The newly developed code is applicable to the detailed safety evaluation necessary to future HTGR design.

Nakagawa, Shigeaki; Saikusa, Akio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Institute (Japan)

2001-02-15T23:59:59.000Z

314

Cooling system for a nuclear reactor  

DOE Patents [OSTI]

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Amtmann, Hans H. (Rancho Santa Fe, CA)

1982-01-01T23:59:59.000Z

315

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building  

SciTech Connect (OSTI)

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

Travis, J.R. [ESSI Inc. (United States); Wilson, T.L.; Spore, J.W.; Lam, K.L. [Los Alamos National Lab., NM (United States); Rao, D.V. [SEA Inc. (United States)

1994-09-01T23:59:59.000Z

316

Cryogenic vessel for a superconducting  

SciTech Connect (OSTI)

A method is described of providing turbocharger compressor outlet pressure to a turbocharger actuator comprising the steps of: forming a circumferential fluid passageway between the compressor housing and an annular ring disposed around the compressor housing; exposing the circumferential fluid passageway to compressor housing pressure; and exposing the turbocharger actuator to the pressure in the circumferential fluid passageway.

Mc Inerney, C.E.

1987-04-07T23:59:59.000Z

317

The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474  

SciTech Connect (OSTI)

The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

2012-07-01T23:59:59.000Z

318

Qualification Requirements of Guided Ultrasonic Waves for Inspection of Piping in Light Water Reactors  

SciTech Connect (OSTI)

Guided ultrasonic waves (GUW) are being increasingly used for both NDT and monitoring of piping. GUW offers advantages over many conventional NDE technologies due to the ability to inspect large volumes of piping components without significant removal of thermal insulation or protective layers. In addition, regions rendered inaccessible to more conventional NDE technologies may be more accessible using GUW techniques. For these reasons, utilities are increasingly considering the use of GUWs for performing the inspection of piping components in nuclear power plants. GUW is a rapidly evolving technology and its usage for inspection of nuclear power plant components requires refinement and qualification to ensure it is able to achieve consistent and acceptable levels of performance. This paper will discuss potential requirements for qualification of GUW techniques for the inspection of piping components in light water reactors (LWRs). The Nuclear Regulatory Commission has adopted ASME Boiler and Pressure Vessel Code requirements in Sections V, III, and XI for nondestructive examination methods, fabrication inspections, and pre-service and in-service inspections. A Section V working group has been formed to place the methodology of GUW into the ASME Boiler and Pressure Vessel Code but no requirements for technique, equipment, or personnel exist in the Code at this time.

Meyer, Ryan M.; Ramuhalli, Pradeep; Doctor, Steven R.; Bond, Leonard J.

2013-08-01T23:59:59.000Z

319

The Gas Reactor Makes a Comeback  

Science Journals Connector (OSTI)

...while the operators of a gas reactor can leave the...ofhis high 699 temperature gas-cooled reactor (HTGR...from the highly pressured turbine drive system might get...would produce combustible gas-es, creating the potential...too much to complete the remaining contracts. So General...

ELIOT MARSHALL

1984-05-18T23:59:59.000Z

320

PIK-2 Reactor with a Low Consumption of High-Enrichment Uranium  

Science Journals Connector (OSTI)

Calculations of the possibility of switching the matrix of the fuel-element cores and the vessel with the reactor envelope to weakly neutron-absorbing aluminum and lowering at the same time the fuel content in...

Yu. V. Petrov; A. N. Erykalov; L. M. Kotova; M. S. Onegin…

2003-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Master external pressure charts  

SciTech Connect (OSTI)

This paper presents a method to develop master external pressure charts from which individual external pressure charts for each material specification may be derived. The master external charts can represent a grouping of materials with similar chemical composition, similar stress-strain curves but produced to different strength levels. External pressure charts are used by various Sections of the ASME Boiler and Pressure Vessel and Piping Codes to design various components such as cylinders, sphered, formed heads, tubes, piping, rings and other components, subjected to external pressure or axial compression loads. These charts are pseudo stress-strain curves for groups of materials with similar stress-strain shapes. The traditional approach was originally developed in the 1940`s and is a graphical approach where slopes to the strain curves are drawn graphically from which pseudo-strain levels are calculated. The new method presented in this paper develops mathematical relationships for the material stress-strain curves and the external pressure charts. The method has the ability to calculate stress-strain curves from existing external pressure charts. The relationships are a function of temperature, the modulus of elasticity, yield strength, and two empirical material constants. In this approach, conservative assumptions used to assign materials to lower bound external pressure charts can be removed. This increases the buckling strength capability of many materials in the Code, providing economic benefits while maintaining the margin of safety specified by the Code criteria. The method can also reduce the number of material charts needed in the Code and provides for the capability to extend the existing pressure charts to higher design temperatures. The new method is shown to contain a number of improvements over the traditional approach and is presently under consideration by appropriate ASME Code committees.

Michalopoulos, E. [Hartford Steam Boiler Inspection and Insurance Co., CT (United States). Codes and Standards Dept.

1996-12-01T23:59:59.000Z

322

Transport reactor development status  

SciTech Connect (OSTI)

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

323

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

324

A case study on effectiveness of structural reliability analysis in nuclear reactor safety assessment  

Science Journals Connector (OSTI)

Problems on reliability of structural integrity occupy an important position in various aspects of nuclear reactor safety. In the present paper, an effective method for quantitative evaluation of structural reliability based on ‘stress strength model’ is developed with the objectives of taking a larger number of factors into the evaluation than before and giving useful results within moderate computing time. The method is applied to the reliability analysis of PWR pressure vessels. The results show the relative importance of inspection as well as the parameter uncertainty for assuring the reliability of the structure, although analysis is limited within the scope of linear elastic fracture mechanics (LEFM). This case study also shows that the analysis of structural reliability is effective for safety assessment of nuclear power plants in general and possibly for the improvements of the consistency in the design code.

A. Yamaguchi; S. Kondo; Y. Togo

1983-01-01T23:59:59.000Z

325

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......inventory (t) 19.8 Fuel Outer diameterpitch...in UO2 No. of fuel assembles 69 Control...3014.0 Reactor vessel Inner diameterheight...of the energy consumption area, aiming...For design of a fuel exchange facility...of the energy consumption area because no...Development of in-vessel type control rod......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

326

Development of Tritium Storage and Transport Vessels  

SciTech Connect (OSTI)

The purpose of this study is to develop tritium storage and transport vessels for industrial applications. Prototype tritium storage and transport vessels were designed and manufactured. Uranium and zirconium/cobalt (ZrCo) metals were selected for the storage materials. The prototype transport container for the vessel was designed on the basis of Type B transportation package standards. The transport container was composed of a steel drum, inner packing materials, and a storage vessel. A second refinement cap was installed on the prototype vessel to protect the valves on the 100 kCi vessel. The vessel is stored in a steel drum packed with a thermal barrier and a shock absorber. Structural, thermal, shielding, and confinement analyses have to be performed for this container based on Type B requirements. (authors)

Paek, S.; Lee, M.; Kim, K.R.; Ahn, D.H. [Korea Atomic Energy Research Institute, Yuseong, Daejeon (Korea, Republic of); Song, K.M.; Shon, S.H. [Korea Electric Power Research Institute, Yuseong-Gu, Daejeon (Korea, Republic of)

2008-07-01T23:59:59.000Z

327

Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

delivery of compressed hydrogen delivery of compressed hydrogen with advanced vessel technology Gene Berry Andrew Weisberg Salvador M. Aceves Lawrence Livermore National Laboratory (925) 422-0864 saceves@LLNL.GOV DOE and FreedomCar & Fuel Partnership Hydrogen Delivery and On-Board Storage Analysis Workshop Washington, DC January 25, 2006 LLNL is developing innovative concepts for efficient containment of hydrogen in light duty vehicles concepts may offer advantages for hydrogen delivery Conformable containers efficiently use available space in the vehicle. We are pursuing multiple approaches to conformability High Strength insulated pressure vessels extend LH 2 dormancy 10x, eliminate boiloff, and enable efficiencies of flexible refueling (compressed/cryogenic H 2 /(L)H 2 ) The PVT properties of H

328

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi nuclear power station  

SciTech Connect (OSTI)

For the decommissioning of the Fukushima-Daiichi Nuclear Power Station (1F), the characterization of fuel-debris in cores of Units 1-3 is necessary. In this study, typical phases of the in-vessel fuel-debris were estimated using a thermodynamic equilibrium (TDE) calculation. The FactSage program and NUCLEA database were applied to estimate the phase equilibria of debris. It was confirmed that the TDE calculation using the database can reproduce the phase separation behavior of debris observed in the Three Mile Island accident. In the TDE calculation of 1F, the oxygen potential [G(O{sub 2})] was assumed to be a variable. At low G(O{sub 2}) where metallic zirconium remains, (U,Zr)O{sub 2}, UO{sub 2}, and ZrO{sub 2} were found as oxides, and oxygen-dispersed Zr, Fe{sub 2}(Zr,U), and Fe{sub 3}UZr{sub 2} were found as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe{sub 3}UZr{sub 2}, decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O{sub 2} is suggested as a typical phase of oxide, and Fe{sub 2}(Zr,U) is suggested as that of metal. However, a more detailed estimation is necessary to consider the distribution of Fe in the reactor pressure vessel through core-melt progression. (authors)

Ikeuchi, Hirotomo; Yano, Kimihiko; Kaji, Naoya; Washiya, Tadahiro [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Kondo, Yoshikazu; Noguchi, Yoshikazu [PESCO Co.Ltd. (Korea, Republic of)

2013-07-01T23:59:59.000Z

329

Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Hydrogen Fuel and Pressure Vessel Forum Hydrogen Fuel and Pressure Vessel Forum The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27-29, 2010 in Beijing, China. High pressure vessel experts gathered to share lessons learned from compressed natural gas (CNG) and hydrogen vehicle deployments, and to identify R&D needs to aid the global harmonization of regulations, codes and standards to enable the successful deployment of hydrogen and fuel cell technologies. The forum also included additional discussion resulting from the DOE and U.S. Department of Transportation (DOT) co-sponsored International Workshop on Compressed Natural Gas and Hydrogen Fuels held on December 10-11, 2009 in Washington, D.C.

330

Plastic instabilities in statically and dynamically loaded spherical vessels  

SciTech Connect (OSTI)

Significant changes were made in design limits for pressurized vessels in the 2007 version of the ASME Code (Section VIII, Div. 3) and 2008 and 2009 Addenda. There is now a local damage-mechanics based strain-exhaustion limit as well as the well-known global plastic collapse limit. Moreover, Code Case 2564 (Section VIII, Div. 3) has recently been approved to address impulsively loaded vessels. It is the purpose of this paper to investigate the plastic collapse limit as it applies to dynamically loaded spherical vessels. Plastic instabilities that could potentially develop in spherical shells under symmetric loading conditions are examined for a variety of plastic constitutive relations. First, a literature survey of both static and dynamic instabilities associated with spherical shells is presented. Then, a general plastic instability condition for spherical shells subjected to displacement controlled and impulsive loading is given. This instability condition is evaluated for six plastic and visco-plastic constitutive relations. The role of strain-rate sensitivity on the instability point is investigated. Calculations for statically and dynamically loaded spherical shells are presented, illustrating the formation of instabilities as well as the role of imperfections. Conclusions of this work are that there are two fundamental types of instabilities associated with failure of spherical shells. In the case of impulsively loaded vessels, where the pulse duration is short compared to the fundamental period of the structure, one instability type is found not to occur in the absence of static internal pressure. Moreover, it is found that the specific role of strain-rate sensitivity on the instability strain depends on the form of the constitutive relation assumed.

Duffey, Thomas A [Los Alamos National Laboratory; Rodriguez, Edward A [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

331

Research reactors - an overview  

SciTech Connect (OSTI)

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.

West, C.D.

1997-03-01T23:59:59.000Z

332

Development of Larger Diameter High Pressure CNG Cylinder Manufactured by Piercing and Drawing for Natural Gas Vehicle  

Broader source: Energy.gov [DOE]

These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 – 29, 2010, in Beijing, China.

333

Self-actuating reactor shutdown system  

DOE Patents [OSTI]

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01T23:59:59.000Z

334

Hanging core support system for a nuclear reactor  

DOE Patents [OSTI]

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Pan, Yen-Cheng (Naperville, IL); Saiveau, James G. (Hickory Hills, IL); Seidensticker, Ralph W. (Wheaton, IL)

1987-01-01T23:59:59.000Z

335

Light Water Reactor Sustainability  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

336

detonation pressure  

Science Journals Connector (OSTI)

detonation pressure ? Detonationsdruck m [Er ist dem Quadrat der Detonationsgeschwindigkeit und der Sprengstoffdichte proportional

2014-08-01T23:59:59.000Z

337

Research Program of a Super Fast Reactor  

SciTech Connect (OSTI)

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)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

338

Microsoft Word - 911103_0_final_rel.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

RPV Reactor Pressure Vessel NGNP Reactor Type Comparison Study Report 9111030 vii SA South Africa TDP Technology Demonstration Program THTR Thorium High Temperature Reactor T...

339

Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A  

SciTech Connect (OSTI)

Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

2009-11-30T23:59:59.000Z

340

MMA Tugboat/ Barge/ Vessel | Open Energy Information  

Open Energy Info (EERE)

MMA Tugboat/ Barge/ Vessel MMA Tugboat/ Barge/ Vessel Jump to: navigation, search Basic Specifications Facility Name MMA Tugboat/ Barge/ Vessel Overseeing Organization Maine Maritime Academy Hydrodynamic Testing Facility Type Tow Vessel Depth(m) 15.2 Water Type Saltwater Cost(per day) Contact POC Special Physical Features Tug: 73 ft (2)16V-92 Detroits Barge: 43 ft by 230ft Research Vessel Friendship: 40 foot vessel w/ 6 cylinder Cummins diesel engine and A-Frame crane Towing Capabilities Towing Capabilities Yes Maximum Velocity(m/s) 5.1 Wavemaking Capabilities Wavemaking Capabilities None Channel/Tunnel/Flume Channel/Tunnel/Flume None Wind Capabilities Wind Capabilities None Control and Data Acquisition Description Full onbard Navigation, GPS, marine radar and depth plotter; standard PC onboard can be configured as needed for data acquisition needs

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

IWTU Construction Workers Set Largest Process Vessel  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

IWTU Construction Workers Set Largest Process Vessel IWTU Construction Workers Set Largest Process Vessel Click on image to enlarge Construction of the Integrated Waste Treatment Unit (IWTU) took a major step forward on Sept. 2, 2009 as crews lifted into place the largest of the six process vessels that will be used to treat radioactive liquid waste stored at the site. The IWTU will use a steam reforming process to solidify the waste for eventual shipment out of Idaho. The vessel and its skid, or framework, were constructed at Premier Technologies in Blackfoot. (Premier is the main small business partner for CH2M-WG Idaho (CWI), the contractor for DOE's Idaho Cleanup Project.) The Carbon Reduction Reformer vessel and skid weigh approximately 60 tons (120,000 lbs.). Because of the weight of the vessel and the location of the

342

Neutron Assay System for Confinement Vessel Disposition  

SciTech Connect (OSTI)

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le}100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Valdez, Jose I. [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

2012-07-13T23:59:59.000Z

343

The Disruption of Vessel-Spanning Bubbles with Sloped Fins in Flat-Bottom and 2:1 Elliptical-Bottom Vessels  

SciTech Connect (OSTI)

Radioactive sludge was generated in the K-East Basin and K-West Basin fuel storage pools at the Hanford Site while irradiated uranium metal fuel elements from the N Reactor were being stored and packaged. The fuel has been removed from the K Basins, and currently, the sludge resides in the KW Basin in large underwater Engineered Containers. The first phase to the Sludge Treatment Project being led by CH2MHILL Plateau Remediation Company (CHPRC) is to retrieve and load the sludge into sludge transport and storage containers (STSCs) and transport the sludge to T Plant for interim storage. The STSCs will be stored inside T Plant cells that are equipped with secondary containment and leak-detection systems. The sludge is composed of a variety of particulate materials and water, including a fraction of reactive uranium metal particles that are a source of hydrogen gas. If a situation occurs where the reactive uranium metal particles settle out at the bottom of a container, previous studies have shown that a vessel-spanning gas layer above the uranium metal particles can develop and can push the overlying layer of sludge upward. The major concern, in addition to the general concern associated with the retention and release of a flammable gas such as hydrogen, is that if a vessel-spanning bubble (VSB) forms in an STSC, it may drive the overlying sludge material to the vents at the top of the container. Then it may be released from the container into the cell’s secondary containment system at T Plant. A previous study demonstrated that sloped walls on vessels, both cylindrical coned-shaped vessels and rectangular vessels with rounded ends, provided an effective approach for disrupting a VSB by creating a release path for gas as a VSB began to rise. Based on the success of sloped-wall vessels, a similar concept is investigated here where a sloped fin is placed inside the vessel to create a release path for gas. A key potential advantage of using a sloped fin compared to a vessel with a sloped wall is that a small fin decreases the volume of a vessel available for sludge storage by a very small fraction compared to a cone-shaped vessel. The purpose of this study is to quantify the capability of sloped fins to disrupt VSBs and to conduct sufficient tests to estimate the performance of fins in full-scale STSCs. Experiments were conducted with a range of fin shapes to determine what slope and width were sufficient to disrupt VSBs. Additional tests were conducted to demonstrate how the fin performance scales with the sludge layer thickness and the sludge strength, density, and vessel diameter based on the gravity yield parameter, which is a dimensionless ratio of the force necessary to yield the sludge to its weight.( ) Further experiments evaluated the difference between vessels with flat and 2:1 elliptical bottoms and a number of different simulants, including the KW container sludge simulant (complete), which was developed to match actual K-Basin sludge. Testing was conducted in 5-in., 10-in., and 23-in.-diameter vessels to quantify how fin performance is impacted by the size of the test vessel. The most significant results for these scale-up tests are the trend in how behavior changes with vessel size and the results from the 23-in. vessel. The key objective in evaluating fin performance is to determine the conditions that minimize the volume of a VSB when disruption occurs because this reduces the potential for material inside the STSC from being released through vents.

Gauglitz, Phillip A.; Buchmiller, William C.; Jenks, Jeromy WJ; Chun, Jaehun; Russell, Renee L.; Schmidt, Andrew J.; Mastor, Michael M.

2010-09-22T23:59:59.000Z

344

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

345

Pressure &Pressure & TemperatureTemperature  

E-Print Network [OSTI]

to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer toprobe to measure atmospheric pressure, and thermometer to measure air temperature.measure air temperature.measure air temperature.measure air temperature

California at Santa Cruz, University of

346

Nuvera fuel cells for Fincantieri marine vessels  

Science Journals Connector (OSTI)

US-based Nuvera Fuel Cells is working with Italian shipbuilder Fincantieri on a programme to power luxury marine vessels with advanced hydrogen PEM fuel cell technology.

2013-01-01T23:59:59.000Z

347

Ordered bed modular reactor design proposal  

SciTech Connect (OSTI)

The Ordered Bed Modular Reactor (OBMR) is a design as an advanced modular HTGR in which the annular reactor core is filled with an ordered bed of fuel spheres. This arrangement allows fuel elements to be poured into the core cavity which is shaped so that an ordered bed is formed and to be discharged from the core through the opening holes in the reactor top. These operations can be performed in a shutdown shorter time. The OBMR has the most of advantages from both the pebble bed reactor and block type reactor. Its core has great structural flexibility and stability, which allow increasing reactor output power and outlet gas temperature as well as decreasing core pressure drop. This paper introduces ordered packing bed characteristics, unloading and loading technique of the fuel spheres and predicted design features of the OBMR. (authors)

Tian, J. [Inst. of Nuclear Energy Technology, Tsinghua Univ., Beijing 100084 (China)

2006-07-01T23:59:59.000Z

348

Compound cryopump for fusion reactors  

E-Print Network [OSTI]

We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

Kovari, M; Shephard, T

2013-01-01T23:59:59.000Z

349

Pressure Safety Program Implementation at ORNL  

SciTech Connect (OSTI)

The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply with pressure safety requirements in 10 CFR 851. It also describes actions taken to develop and implement ORNL’s Pressure Safety Program.

Lower, Mark [ORNL; Etheridge, Tom [ORNL; Oland, C. Barry [XCEL Engineering, Inc.

2013-01-01T23:59:59.000Z

350

Device for inspecting vessel surfaces  

DOE Patents [OSTI]

A portable, remotely-controlled inspection crawler for use along the walls of tanks, vessels, piping and the like. The crawler can be configured to use a vacuum chamber for supporting itself on the inspected surface by suction or a plurality of magnetic wheels for moving the crawler along the inspected surface. The crawler is adapted to be equipped with an ultrasonic probe for mapping the structural integrity or other characteristics of the surface being inspected. Navigation of the crawler is achieved by triangulation techniques between a signal transmitter on the crawler and a pair of microphones attached to a fixed, remote location, such as the crawler's deployment unit. The necessary communications are established between the crawler and computers external to the inspection environment for position control and storage and/or monitoring of data acquisition.

Appel, D. Keith (Aiken, SC)

1995-01-01T23:59:59.000Z

351

V1.6 Development of Advanced Manufacturing Technologies for Low Cost Hydrogen Storage Vessels  

SciTech Connect (OSTI)

The goal of this project is to develop an innovative manufacturing process for Type IV high-pressure hydrogen storage vessels, with the intent to significantly lower manufacturing costs. Part of the development is to integrate the features of high precision AFP and commercial FW. Evaluation of an alternative fiber to replace a portion of the baseline fiber will help to reduce costs further.

Leavitt, Mark; Lam, Patrick; Nelson, Karl M.; johnson, Brice A.; Johnson, Kenneth I.; Alvine, Kyle J.; Ruiz, Antonio; Adams, Jesse

2012-10-01T23:59:59.000Z

352

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

Broader source: Energy.gov (indexed) [DOE]

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

353

Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications  

SciTech Connect (OSTI)

Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core. Still there are problems of containment since many of the proposed vessel materials such as W or Mo have high neutron cross sections making the design of a critical system difficult. There is also the possibility for a GCR to remain in a subcritical state, and by the use of a shockwave mechanism, increase the pressure and temperature inside the core to achieve criticality. This type of GCR is referred to as a shockwave-driven pulsed gas core reactor. These two basic designs were evaluated as advance concepts for space power and propulsion.

Samim Anghaie

2002-08-13T23:59:59.000Z

354

Foam vessel for cryogenic fluid storage  

DOE Patents [OSTI]

Cryogenic storage and separator vessels made of polyolefin foams are disclosed, as are methods of storing and separating cryogenic fluids and fluid mixtures using these vessels. In one embodiment, the polyolefin foams may be cross-linked, closed-cell polyethylene foams with a density of from about 2 pounds per cubic foot to a density of about 4 pounds per cubic foot.

Spear, Jonathan D (San Francisco, CA)

2011-07-05T23:59:59.000Z

355

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect (OSTI)

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab.

Silady, F.A.; Millunzi, A.C.

1989-08-01T23:59:59.000Z

356

Westinghouse Small Modular Reactor balance of plant and supporting systems design  

SciTech Connect (OSTI)

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01T23:59:59.000Z

357

Uncertainty Analysis for a De-pressurised Loss of Forced Cooling Event of the PBMR Reactor  

SciTech Connect (OSTI)

This paper presents an uncertainty analysis for a De-pressurised Loss of Forced Cooling (DLOFC) event that was performed with the systems CFD (Computational Fluid Dynamics) code Flownex for the PBMR reactor. An uncertainty analysis was performed to determine the variation in maximum fuel, core barrel and reactor pressure vessel (RPV) temperature due to variations in model input parameters. Some of the input parameters that were varied are: thermo-physical properties of helium and the various solid materials, decay heat, neutron and gamma heating, pebble bed pressure loss, pebble bed Nusselt number and pebble bed bypass flows. The Flownex model of the PBMR reactor is a 2-dimensional axisymmetrical model. It is simplified in terms of geometry and some other input values. However, it is believed that the model adequately indicates the effect of changes in certain input parameters on the fuel temperature and other components during a DLOFC event. Firstly, a sensitivity study was performed where input variables were varied individually according to predefined uncertainty ranges and the results were sorted according to the effect on maximum fuel temperature. In the sensitivity study, only seven variables had a significant effect on the maximum fuel temperature (greater that 5 deg. C). The most significant are power distribution profile, decay heat, reflector properties and effective pebble bed conductivity. Secondly, Monte Carlo analyses were performed in which twenty variables were varied simultaneously within predefined uncertainty ranges. For a one-tailed 95% confidence level, the conservatism that should be added to the best estimate calculation of the maximum fuel temperature for a DLOFC was determined as 53 deg. C. This value will probably increase after some model refinements in the future. Flownex was found to be a valuable tool for uncertainly analyses, facilitating both sensitivity studies and Monte Carlo analyses. (authors)

Jansen van Rensburg, Pieter A.; Sage, Martin G. [PBMR, 1279 Mike Crawford Avenue, Centurion 0046 (South Africa)

2006-07-01T23:59:59.000Z

358

Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project  

SciTech Connect (OSTI)

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

A. B. Culp

2007-01-26T23:59:59.000Z

359

Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor  

SciTech Connect (OSTI)

The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (authors)

Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia [Department of Nuclear Engineering, Politecnico di Milano, Via Ponzio, 34/3, 20133 Milano (Italy)

2004-07-01T23:59:59.000Z

360

Light Water Reactors Technology Development - Nuclear Reactors  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995  

SciTech Connect (OSTI)

The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

NONE

1998-09-01T23:59:59.000Z

362

NETL: Pressure Swing Absorption Device  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Pressure Swing Absorption Device and Process for Separating CO2 from Shifted Syngas and its Capture for Subsequent Storage Pressure Swing Absorption Device and Process for Separating CO2 from Shifted Syngas and its Capture for Subsequent Storage Project No.: DE-FE0001323 New Jersey Institute of Technology is developing an advanced pressure swing absorption-based (PSAB) device via laboratory-based experiments. The device will be used to accomplish a cyclic process to process low temperature post-shift-reactor synthesis gas resulting from the gasification process into purified hydrogen at high pressure for use by the combustion turbine of an integrated gasification combined cycle (IGCC) plant. The overall goal of the proposed work is to develop an advanced PSAB device and cyclic process for use in a coal-fired IGCC plant to produce purified hydrogen at high pressure and a highly purified CO2 stream suitable for use or sequestration.

363

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report  

SciTech Connect (OSTI)

The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

364

E-Print Network 3.0 - adjacent vessel sign Sample Search Results  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

three-dimensional (3-D) data, which are vessel voxel projection probability, vessel detection... probability, false vessel probability, and vessel-tissue contrast-to-noise ratio...

365

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect (OSTI)

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

366

Thermal wake/vessel detection technique  

DOE Patents [OSTI]

A computer-automated method for detecting a vessel in water based on an image of a portion of Earth includes generating a thermal anomaly mask. The thermal anomaly mask flags each pixel of the image initially deemed to be a wake pixel based on a comparison of a thermal value of each pixel against other thermal values of other pixels localized about each pixel. Contiguous pixels flagged by the thermal anomaly mask are grouped into pixel clusters. A shape of each of the pixel clusters is analyzed to determine whether each of the pixel clusters represents a possible vessel detection event. The possible vessel detection events are represented visually within the image.

Roskovensky, John K. (Albuquerque, NM); Nandy, Prabal (Albuquerque, NM); Post, Brian N (Albuquerque, NM)

2012-01-10T23:59:59.000Z

367

The Results From the First High-Pressure Melt Ejection Test Completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories  

SciTech Connect (OSTI)

A high-pressure melt ejection test using prototypical corium was conducted at Atomic Energy of Canada Limited Chalk River Laboratories. This test was planned by the CANDU Owners Group to study the potential for an energetic interaction between molten fuel and water under postulated single-channel flow-blockage events. The experiments were designed to address regulator concerns surrounding this very low probability postulated accident events in CANDU Pressurized Heavy Water Reactors. The objective of the experimental program is to determine whether a highly energetic 'steam explosion' and associated high-pressure pulse, is possible when molten material is finely fragmented as it is ejected from a fuel channel into the heavy-water moderator. The finely fragmented melt particles would transfer energy to the moderator as it is dispersed, creating a modest pressure pulse in the calandria vessel. The high-pressure melt ejection test consisted of heating up a {approx} 5 kg thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3} inside a 1.14-m length of insulated pressure tube. When the molten material reached the desired temperature of {approx} 2400 deg C, the pressure inside the tube was raised to 11.6 MPa, failing the pressure tube at a pre-machined flaw, and releasing the molten material into the surrounding tank of 68 deg C water. The experiment investigated the dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels. The measured mean particle size was 0.686 mm and the peak dynamic pressures were between 2.54 and 4.36 MPa, indicating that an energetic interaction between the melt and the water did not occur in the test. (authors)

Nitheanandan, T.; Kyle, G.; O'Connor, R.; Sanderson, DB. [Chalk River Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario, Canada, K0J 1J0 (Canada)

2006-07-01T23:59:59.000Z

368

Metrology/viewing system for next generation fusion reactors  

SciTech Connect (OSTI)

Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system.

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M. [Oak Ridge National Lab., TN (United States); Dagher, M.A. [Boeing Rocketdyne Div., Canoga Park, CA (United States)

1997-02-01T23:59:59.000Z

369

High flux isotope reactor: Quarterly report October through December 1986  

SciTech Connect (OSTI)

Two routine cycles of operation of the HFIR reactor were completed during the quarter. The shutdowns to end these cycles were both scheduled. The end-of-cycle 287 shutdown was extended indefinitely to investigate the embrittlement of reactor vessel materials due to radiation damage. The reactor remains down at the end of the quarter. Following the scheduled end-of-cycle 287 shutdown period, subsequent shutdown time was designated as unscheduled. The two scheduled shutdowns, fourth quarter downtime resulting from a third quarter scheduled shutdown, and the extended unscheduled shutdown account for the low 44.2% on-stream time for the quarter. The scheduled control plate replacement and vessel internals inspection was completed at the end-of-cycle 287. The inspection revealed a blister on control cylinder 9. This flaw was attributed to a manufacturing defect.

Corbett, B.L.; Farrar, M.B.

1987-04-01T23:59:59.000Z

370

Photocatalytic reactor  

DOE Patents [OSTI]

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

1999-01-19T23:59:59.000Z

371

High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant  

SciTech Connect (OSTI)

The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

J. M. Beck; L. F. Pincock

2011-04-01T23:59:59.000Z

372

Final Vitrification Melter And Vessels Evaluation Documentation  

Broader source: Energy.gov [DOE]

DOE has prepared final evaluations and made waste incidental to reprocessing determinations for the vitrification melter and feed vessels (the concentrator feed makeup tank and the melter feed hold...

373

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

374

Osmotic Pressure  

Science Journals Connector (OSTI)

... Whetham in which he attempts to consign actual experimental work on osmotic pressure to the humble rôle of showing how far the assumptions made in so-called thermodynamical proofs can be ... actual osmotic processes, and that the experimental work on osmotic pressure does not play that humble rôle to which Mr. Whetham would consign it. ...

LOUIS KAHLENBERG

1906-07-05T23:59:59.000Z

375

ASME post construction pressure technology codes  

SciTech Connect (OSTI)

The need to continue to operate pressurized equipment and other facilities in a safe, reliable and cost effective manner has led to the development of many new approaches to in-service inspection, flaw evaluation, and repair. Interest on the part of users, regulatory authorities and others in standardizing these approaches has led to the formation of a new ASME Main Committee on Post Construction under the Board on Pressure Technology Codes and Standards, and a new Division of the Pressure Vessel Research Council on Continued Operation of Equipment. This paper provides a brief overview of these activities.

Sims, J.R. [Exxon Research and Engineering Co., Florham Park, NJ (United States)

1996-12-01T23:59:59.000Z

376

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

377

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

378

Routine and post-accident sampling in nuclear reactors  

SciTech Connect (OSTI)

Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experience.

Armento, W.J.; Kitts, F.G.; German, G.E.

1980-01-01T23:59:59.000Z

379

The next generation of power reactors - safety characteristics  

SciTech Connect (OSTI)

The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs.

Modro, S.M.

1995-01-01T23:59:59.000Z

380

Pressurized Items and Cryogens Assessment plan - Developed By NNSA/Nevada Site Office Facility Representative Division  

Broader source: Energy.gov (indexed) [DOE]

Pressurized Systems and Cryogens Pressurized Systems and Cryogens Performance Objective: Assure personnel health and safety through regularly scheduled inspections and maintenance on pressure vessels and equipment, compressed gases and gas cylinders, vacuum equipment and systems, hydraulics, and cryogenic materials and systems. Performance Criteria: BN inspects, operates and safely stores unmodified compressed-gas or liquid cylinders approved by the Department of Transportation (DOT) and the appropriate regulators. BN inspects, operates and maintains refrigeration systems that comply with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Codes, and applicable Air Conditioning and Refrigeration Institute (ARI) standards. BN inspects, operates and maintains pressure systems that operate at an

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Simulation of the Fuel Reactor of a Coal?Fired Chemical Looping Combustor  

Science Journals Connector (OSTI)

Responsible carbon management (CM) will be required for the future utilization of coal for power generation. CO 2 separation is the more costly component of CM not sequestration. Most methods of capture require a costly process of gas separation to obtain a CO 2 ?rich gas stream. However recently a process termed Chemical Looping Combustion (CLC) has been proposed in which an oxygen?carrier is used to provide the oxygen for combustion. This process quite naturally generates a separate exhaust gas stream containing mainly H 2 O and CO 2 but requires two reaction vessels an Air Reactor (AR) and a Fuel Reactor (FR). The carrier (M for metal the usual carrier) is oxidized in the AR. This highly exothermic process provides heat for power generation. The oxidized carrier (MO) is separated from this hot vitiated air stream and transported to the FR where it oxidizes the hydrocarbon fuel yielding an exhaust gas stream of mainly H 2 O and CO 2 . This process is usually slightly endothermic so that the carrier must also transport the necessary heat of reaction. The reduced carrier (M) is then returned to the air reactor for regeneration hence the term “looping.” The net chemical reaction and energy release is identical to that of conventional combustion of the fuel. However CO 2 separation is easily achieved the only operational penalty being the slight pressure losses required to circulate the carrier. CLC requires many unit operations involving gas?solid or granular flow. To utilize coal in the fuel reactor in either a moving bed or bubbling fluidized bed the granular flow is especially critical. The solid coal fuel must be heated by the recycled metal oxide driving off moisture and volatile material. The remaining char must be gasified by H 2 O (or CO 2 ) which is recycled from the product stream. The gaseous product of these reactions must then contact the MO before leaving the bed to obtain complete conversion to H 2 O and CO 2 . Further the reduced M particles must be removed from the bed and returned to the air reactor without any accompanying unburned fuel. This paper presents a simulation of the gas?particle granular flow with heat transfer and chemical reactions in the FR. Accurate simulation of the segregation processes depending on particle density and size differences between the carrier and the fuel allows the design of a reactor with the desired behavior.

Kartikeya Mahalatkar; Thomas O’Brien; E. David Huckaby; John Kuhlman

2009-01-01T23:59:59.000Z

382

Development of Novel Water-Gas-Shift Membrane Reactor  

E-Print Network [OSTI]

Development of Novel Water- Gas-Shift Membrane Reactor Addressing Barrier L: H2 Purification-22, 2003 #12;Water-Gas-Shift Membrane Reactor · Relevance/Objectives - Produce Enhanced H2 Product with ppm CO at High Pressure Used for Reforming - Overcome Barrier L: H2 Purification/CO Clean-up - Achieve

383

Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up  

SciTech Connect (OSTI)

A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked 'solver'.

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University 71348-51154, Shiraz (Iran, Islamic Republic of)

2009-09-09T23:59:59.000Z

384

Large Scale Weather Control Using Nuclear Reactors  

E-Print Network [OSTI]

It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

Moninder Singh Modgil

2002-10-02T23:59:59.000Z

385

Particle Sizing using Passive Ultrasonic Measurement of Vessel Wall Vibrations  

E-Print Network [OSTI]

Particle Sizing using Passive Ultrasonic Measurement of Vessel Wall Vibrations Gillian Carson for particle sizing using an ultrasonic transducer to measure vessel wall vibrations and 1 #12;considers in a stirred vessel, its subse- quent impact with the vessel wall, and the resulting flexural vibrations

Mottram, Nigel

386

Heat exchanger for reactor core and the like  

DOE Patents [OSTI]

A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

Kaufman, Jay S. (Del Mar, CA); Kissinger, John A. (Del Mar, CA)

1986-01-01T23:59:59.000Z

387

Thermal Behavior of SST-1 Vacuum Vessel and Plasma Facing Components during Baking  

Science Journals Connector (OSTI)

Abstract For plasma physics experiment, the baking of vacuum vessel (VV) as well as plasma facing components (PFC) of Steady-state Superconducting Tokamak (SST-1) is very essential. SST-1 vacuum vessel consists of ultra-high vacuum (UHV) compatible sixteen sectors in which U-shaped baking channels are embedded on inner surfaces of each of them. Similarly, \\{PFCs\\} are comprised of modular graphite diverters and movable graphite based limiters and stainless steel (SS 304L) tubes are brazed on the back plate of PFC for baking. Baking of SST-1 vacuum vessel and plasma facing components are carried out using nitrogen gas heating and supply system. SST-1 main vacuum vessel is baked at 150 °C by circulating hot nitrogen gas at 250 °C at 4.5 bar gauge (g) pressure through these U-shaped channels. The plasma facing components (PFC) are baked at 250 °C or more in the similar fashion by passing hot nitrogen gas through these SS brazed tubes. Thermal analysis shows that the temperature of 150 °C at the vacuum vessel is achieved within ten hours if hot nitrogen gas is passed at the ramp rate of 50 °C/h while thermal shields are maintained at 85 K. It is also observed that the baking of either of them at a given temperature could be possible through radiation if one of them is maintained at desired temperature. The vacuum vessel at room temperature could be baked to 150 °C due to radiation from PFC after 40 hours when PFC alone is baked at 150 °C. The mass flow rate required to bake SST-1 vacuum vessel at 150 °C is 1.074 kg/s while that for raising \\{PFCs\\} temperature to 150 °C is 0.57 kg/s. The mass flow rate required to bake \\{PFCs\\} at 250 °C is 0.80 kg/s.

Ziauddin Khan; Yuvakiran Paravastu; Subrata Pradhan

2014-01-01T23:59:59.000Z

388

Miniature Specimen Testing The Small Punch Test (SPT) has successfully been used in the US and in Japan  

E-Print Network [OSTI]

and in Japan to assess the material behavior of steels in nuclear reactor and pressure vessels as well

Berlin,Technische Universität

389

Numerical investigation of the convection of heat-emitting liquid reactor materials taking stratification into account  

SciTech Connect (OSTI)

Convective heat transfer to a reactor vessel following core destruction is analyzed. Fuel fragment and structural materials are assumed to melt either mix uniformly or stratify. The Navier-Stokes equations were solved numerically with a rectangular coarse difference grid, neglecting the negligibly small contribution of the laminar boundary layer. Calculations for different levels of volume heat release showed that integral heat fluxes at the lateral and top surfaces of the melt are virtually independent of the convective flow scenario. However, the maximum heat flux on the lateral surface is approximately 1.5 times higher for the homogeneous case than for the stratified case. The higher heat flux could result in larger mechanical loads on the reactor vessel, requiring more cooling of the reactor vessel outer surface. 12 refs., 4 figs.

Likhanskii, V.V.; Loboiko, A.I.; Khoruzhii, O.V.

1994-11-01T23:59:59.000Z

390

SRS Small Modular Reactors  

SciTech Connect (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2012-04-27T23:59:59.000Z

391

SRS Small Modular Reactors  

ScienceCinema (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2014-05-21T23:59:59.000Z

392

Review of High Temperature Water and Steam Cooled Reactor Concepts  

SciTech Connect (OSTI)

This review summarizes design concepts of supercritical-pressure water cooled reactors (SCR), nuclear superheaters and steam cooled fast reactors from 1950's to the present time. It includes water moderated supercritical steam cooled reactor, SCOTT-R and SC-PWR of Westinghouse, heavy water moderated light water cooled SCR of GE, SCLWR and SCFR of the University of Tokyo, B-500SKDI of Kurchatov Institute, CANDU -X of AECL, nuclear superheaters of GE, subcritical-pressure steam cooled FBR of KFK and B and W, Supercritical-pressure steam cooled FBR of B and W, subcritical-pressure steam cooled high converter by Edlund and Schultz and subcritical-pressure water-steam cooled FBR by Alekseev. This paper is prepared based on the previous review of SCR2000 symposium, and some author's comments are added. (author)

Oka, Yoshiaki [Nuclear Engineering Research Laboratory, The University of Tokyo, 3-1, Hongo 7-Chome, Bunkyo-ku (Japan)

2002-07-01T23:59:59.000Z

393

In-vessel coolability and retention of a core melt. Volume 2  

SciTech Connect (OSTI)

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

1996-10-01T23:59:59.000Z

394

Reactor cooldown time cut 30-50% with liquid nitrogen injection  

SciTech Connect (OSTI)

Cooling the hydrotreater in preparation for a catalyst bed turnaround was costly and time consuming for Champlin Petroleum Company's Corpus Christi, TX, refinery. The hydrotreater (reactor) functions to upgrade sour gas oil feedstock in a high pressure (1200 psig) catalytic reaction with hydrogen. When the catalyst becomes spent, the 4 1/2'' thick steel reactor and internals have to be cooled down from 700-750/sup 0/F operating temperature to a temperature that is safe for personnel to enter the vessel. During the time the hydotreater is out of service, the refinery must switch to a more expensive sweet crude feedstock. Cooldown was taking 24-36 hours just to reach 120-140/sup 0/F, about the maximum temperature acceptable for personnel entry. Champlin Petroleum Company evaluated the economics and decided to use liquid nitrogen to accelerate the cooldown of the hydrotreater. A company which offered the liquid nitrogen cooling service provided technical backup to accurately estimate the mass of material that had to be cooled and the costs associated with options when the nitrogen could be introduced during the cooldown cycle. Champlin decided to buy the liquid nitrogen injection as a package service. With the cooldown accelerated by using liquid nitrogen, it takes only 18 hours to cool the reactor to 90/sup 0/F compared to the 24-36 hours it used to take to reach 120-140/sup 0/F. Not only is significant cooldown time savings realized, but the catalyst is discharged at the cooler 90/sup 0/F, making handling easier and safer.

Cornelius, R.; Hodel, A.E.

1985-08-01T23:59:59.000Z

395

Linear chain tensioning of moored production vessels  

SciTech Connect (OSTI)

Part 1 of this two-part series discussed the worldwide floating production vessel (FPV) market and evolution of the linear puller concept. The three principal types of chain jack systems - hollow ram, single and twin cylinders - were introduced. And advantages of this relatively new form of passive mooring were outlined. This concluding article covers applications of linear chain pullers on various vessels, including use on an example 35,000-t North Sea semi-submersible. Chain wear and how linear pullers avoid wear associated with windlass-type systems are discussed, along with the optimization possible through use of a swiveling chain fair-lead latch (SCFL).

Peters, B. (Bardex Corp., London (United Kingdom))

1993-06-01T23:59:59.000Z

396

Nuclear reactor alignment plate configuration  

DOE Patents [OSTI]

An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

2014-01-28T23:59:59.000Z

397

Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

398

Monitoring system for a liquid-cooled nuclear fission reactor. [PWR  

DOE Patents [OSTI]

The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

DeVolpi, A.

1984-07-20T23:59:59.000Z

399

Fluidized bed steam reactor including two horizontal cyclone separators and an integral recycle heat exchanger  

SciTech Connect (OSTI)

A reactor is described comprising: a vessel; a first furnace section disposed in said vessel; a second furnace section disposed in said vessel; means in each of said furnace sections for receiving a combustible fuel for generating heat and combustion gases; a first heat recovery area located adjacent said furnace sections; a second heat recovery area located adjacent said furnace sections; means for passing said combustion gases from said first furnace section to said first heat recovery area; and means for passing said combustion gases from said second furnace section to said second heat recovery area.

Gorzegno, W.P.

1993-06-15T23:59:59.000Z

400

Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System  

SciTech Connect (OSTI)

The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-epsilon model with two-layer all y+ wall treatment performs better than the other k-epsilon and k-omega turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results. A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.

Angelo Frisani; Yassin A. Hassan; Victor M. Ugaz

2010-11-02T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Numerical tools applied to power reactor noise analysis  

SciTech Connect (OSTI)

In this paper, the development of numerical tools allowing the determination of the neutron noise in power reactors is reported. These tools give the space-dependence of the fluctuations of the neutron flux induced by fluctuating properties of the medium in the 2-group diffusion approximation and in a 2-dimensional representation of heterogeneous systems. Some applications of these tools to power reactor noise analysis are then described. These applications include the unfolding of the noise source from the resulting neutron noise measured at a few discrete locations throughout the core, the study of the space-dependence of the Decay Ratio in Boiling Water Reactors, the noise-based estimation of the Moderator Temperature Coefficient of reactivity in Pressurized Water Reactors, the modeling of shell-mode core barrel vibrations in Pressurized Water Reactors, and the investigation of the validity of the point-kinetic approximation in subcritical systems. (authors)

Demaziere, C.; Pazsit, I. [Chalmers Univ. of Technology, Dept. of Nuclear Engineering, SE-412 96 Goeteborg (Sweden)

2006-07-01T23:59:59.000Z

402

Microsoft Word - Milestone Report-Assessment of Surveillance...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

ORNLLTR-2011172 Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials June 2011 Prepared by R.K. Nanstad,...

403

Generating Unstructured Nuclear Reactor Core Meshes in Parallel  

Science Journals Connector (OSTI)

Abstract Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

Rajeev Jain; Timothy J. Tautges

2014-01-01T23:59:59.000Z

404

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning  

Broader source: Energy.gov (indexed) [DOE]

American Recovery and Reinvestment American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Sa- vannah River Site this month after they safely re- moved its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the work- ers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Workers will cut the dome into smaller pieces for disposal. Removal of the dome allows workers to access the 219,000-pound reactor vessel and two steam generators so they can remove and permanently dispose them onsite. Re- maining equipment will be moved to the cavity vacated by the vessel, and below-grade portions of the reactor will be

405

An autonomous long-term fast reactor system and the principal design limitations of the concept  

E-Print Network [OSTI]

Actinides MOX Mixed OXide MSR Molten-Salt Reactors NERI Nuclear Energy Research Initiative vii PWR Pressurized Water Reactor RGPu Reactor-Grade Plutonium SCNES Self-Consistent Nuclear Energy System STAR Secure Transportable Autonomous Reactor... of LWR?s, the drastic increase of Am and Cm inventories are observed after uranium fuel irradiation and the second recycling of MOX fuel.1 Therefore, partitioning and transmutation of the recovered MA?s could significantly reduce the long...

Tsvetkova, Galina Valeryevna

2004-09-30T23:59:59.000Z

406

Ensemble Classification System Applied for Retinal Vessel Segmentation on Child Images Containing Various Vessel Profiles  

Science Journals Connector (OSTI)

This paper describes a new supervised method for segmentation of blood vessels in retinal images of multi ethnic school children. This method uses an ensemble classification system of boot strapped decision tr...

M. M. Fraz; P. Remagnino; A. Hoppe; B. Uyyanonvara…

2012-01-01T23:59:59.000Z

407

A Laser Metrology/Viewing System for ITER In-Vessel Inspection  

SciTech Connect (OSTI)

This paper identifies the requirements for a remotely operated precision laser ranging system for the International Thermonuclear Experimental Reactor. The inspection system is used for metrology and viewing, and must be capable of achieving submillimeter accuracy and operation in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field levels. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser-optic module linked through fiberoptics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic mast. Gamma irradiation up to 10{sup 7} Gy was conducted on critical sensor components with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway.

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M. [Oak Ridge National Lab., TN (United States); Dagher, M.A. [Boeing Rocketdyne Div., Canoga Park, CA (United States); Slotwinski, A. [Coleman Research Corp., Springfield, VA (United States)

1997-12-31T23:59:59.000Z

408

Pressure, temperature, and dissolved gas dependence of dielectric breakdown in water.  

Science Journals Connector (OSTI)

It has been shown experimentally that the optical breakdown strength of water is a pressure dependent quantity growing with increasing pressure. The dependence of the breakdown strength on temperature and dissolved gas concentration over a larger range of pressures will be observed. Using a custom fabricated pressure vessel and high?power Nd:YAG laser breakdown events will be generated and observed over a range of pressures from 0 to 25 kpsi. Observations of breakdown events will be made using a high?speed photodetector located behind the pressure vessel’s optical windows. Dissolved gas concentration will be controlled and varied using a custom water preparation system over a range from water’s vapor pressure (?20 torr) to atmospheric pressure.Temperature will be monitored using a thermocouple attached to the pressure vessel and the temperature dependence will be measured over a range from 20 to 35 °C. A comparison between current single detector methods and previous imaging methods of using breakdown to determine absolute pressure will then be made. [Work supported by Impulse Devices Inc.

Jonathan Sukovich; R. Glynn Holt

2010-01-01T23:59:59.000Z

409

Study Reveals Challenges and Opportunities Related to Vessels...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Study Reveals Challenges and Opportunities Related to Vessels for U.S. Offshore Wind Study Reveals Challenges and Opportunities Related to Vessels for U.S. Offshore Wind October 1,...

410

Microsoft PowerPoint - Mod 10a - Vessel System - final.ppt [Compatibil...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

(for residual heat removal) * Heat exchanger vessels insulated (to minimize parasitic heat loss) 6 Outline * Vessel System functions and requirements Vessel System functions...

411

Thermal Analysis to Calculate the Vessel Temperature and Stress in Alcator C-Mod Due to the Divertor Upgrade  

SciTech Connect (OSTI)

Alcator C-Mod is planning an upgrade to its outer divertor. The upgrade is intended to correct the existing outer divertor alignment with the plasma, and to operate at elevated temperatures. Higher temperature operation will allow study of edge physics behavior at reactor relevant temperatures. The outer divertor and tiles will be capable of operating at 600oC. Longer pulse length, together with the plasma and RF heat of 9MW, and the inclusion of heater elements within the outer divertor produces radiative energy which makes the sustained operation much more difficult than before. An ANSYS model based on ref. 1 was built for the global thermal analysis of C-Mod. It models the radiative surfaces inside the vessel and between the components, and also includes plasma energy deposition. Different geometries have been simulated and compared. Results show that steady state operation with the divertor at 600oC is possible with no damage to major vessel internal components. The differential temperature between inner divertor structure, or "girdle" and inner vessel wall is ~70oC. This differential temperature is limited by the capacity of the studs that hold the inner divertor backing plates to the vessel wall. At a 70oC temperature differential the stress on the studs is within allowable limits. The thermal model was then used for a stress pass to quantify vessel shell stresses where thermal gradients are significant.

Han Zhang, Peter H. Titus, Robert Ellis, Soren Harrison and Rui Vieira

2012-08-29T23:59:59.000Z

412

Pressure transducer  

DOE Patents [OSTI]

A pressure transducer suitable for use in high temperature environments includes two pairs of induction coils, each pair being bifilarly wound together, and each pair of coils connected as opposite arms of a four arm circuit; an electrically conductive target moveably positioned between the coil pairs and connected to a diaphragm such that deflection of the diaphragm causes axial movement of the target and an unbalance in the bridge output.

Anderson, Thomas T. (Downers Grove, IL); Roop, Conard J. (Lockport, IL); Schmidt, Kenneth J. (Midlothian, IL); Gunchin, Elmer R. (Lockport, IL)

1989-01-01T23:59:59.000Z

413

Probabilistic risk assessment of N Reactor using NUREG-1150 methods  

SciTech Connect (OSTI)

A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a US Department of Energy (DOE) Category A production reactor. The main contractor is Westinghouse Hanford Company (Westinghouse Hanford). The PRA methodology developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL) in support of the NUREG-1150 (Reference 1) effort were used for this analysis. N Reactor is a graphite-moderated pressurized water reactor designed by General Electric. The dual-purpose 4000 MWt nuclear plant is located within the Hanford Site in the south-central part of the State of Washington. In addition to producing special materials for the DOE, N Reactor generates 860 MWe for the Washington Public Power Supply System. The reactor has been operated successfully and safely since 1963, and was put into standby status in 1988 due to the changing need in special nuclear material. 3 refs., 4 tabs.

Wang, O.S.; Baxter, J.T.; Coles, G.A.; Powers, T.B.; Zentner, M.D.

1989-11-01T23:59:59.000Z

414

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

415

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

416

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

417

Public comment sought on final end state of Experimental Breeder Reactor-II  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Media Contacts: Danielle Miller, 208-526-5709 Media Contacts: Danielle Miller, 208-526-5709 Joseph Campbell, CWI, 208-360-0142 Public comment sought on final end state of Experimental Breeder Reactor-II The U.S. Department of Energy (DOE) is seeking public comment on a range of alternatives for disposition of the landmark Experimental Breeder Reactor-II (EBR-II) building and reactor vessel at the Idaho Site's Materials and Fuels Complex. An Engineering Evaluation/Cost Analysis (EE/CA) document with four proposed alternatives for the final end state of the reactor facility and support structures is currently under evaluation by DOE, the U.S. Environmental Protection Agency, and Idaho's Department of Environmental Quality. Experimental Breeder Reactor-II containment dome The EBR-II was an innovative sodium-cooled reactor with an output of 62

418

Blood vessel segmentation methodologies in retinal images - A survey  

Science Journals Connector (OSTI)

Retinal vessel segmentation algorithms are a fundamental component of automatic retinal disease screening systems. This work examines the blood vessel segmentation methodologies in two dimensional retinal images acquired from a fundus camera and a survey ... Keywords: Blood vessel segmentation, Image segmentation, Medical imaging, Retinal images, Retinopathy, Survey

M. M. Fraz; P. Remagnino; A. Hoppe; B. Uyyanonvara; A. R. Rudnicka; C. G. Owen; S. A. Barman

2012-10-01T23:59:59.000Z

419

Ship-owners' decisions to outsource vessel management  

E-Print Network [OSTI]

of domiciliation, number of vessels). In addition, a specific country effect is identified for Greek shipEA 4272 Ship-owners' decisions to outsource vessel management Pierre Cariou* Francois-Charles Wolff,version1-17May2011 #12;Ship-owners' decisions to outsource vessel management Pierre CARIOU Corresponding

Paris-Sud XI, Université de

420

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Improved vortex reactor system  

DOE Patents [OSTI]

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

422

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28T23:59:59.000Z

423

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect (OSTI)

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

424

The Argonaut Reactor - Reactors designed/built by Argonne National  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

425

Life cycle considerations in propulsion alternatives for fast vessels  

SciTech Connect (OSTI)

Fast vessels are being built and operated for a large range of passenger-carrying applications. Fast cargo-carrying vessels are being considered in a variety of sizes as well. A major decision in design and construction of these vessels is the propulsion system; this decision has major impacts on the operation economics as well as the operational capabilities of the vessels. Factors involved in consideration of propulsion alternatives for fast vessels are examined, with emphasis upon the total life cycle operating implications of these factors. A methodology for considering the factors is suggested, and an example is presented with results of the consideration tradeoffs.

Luck, D.L. [General Electric Co., Evendale, OH (United States). GE Marine and Industrial Engines

1996-07-01T23:59:59.000Z

426

Confinement Vessel Assay System: Calibration and Certification Report  

SciTech Connect (OSTI)

Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of SNM in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of {le} 100-g {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

2012-07-17T23:59:59.000Z

427

REACTOR REFUELING - INTERIM DECAY STORAGE (FFTF)  

SciTech Connect (OSTI)

The IDS facility is located between the CLEM rails and within the FFTF containment building. It is located in a rectangular steel-lined concrete cell which lies entirely below the 550 ft floor level with the top flush with the 550 ft floor level. The BLTC rails within containment traverse the IDS cover (H-4-38001). The facility consists of a rotatable storage basket submerged in liquid sodium which is contained in a stainless steel tank. The storage positions within the basket are arranged so that it is not physically possible to achieve a critical array. The primary vessel is enclosed in a secondary guard tank of such size and arrangement that, should a leak develop in the primary tank, the sodium level would not fall below the top of the fueled section of the stored core components or test assemblies. The atmosphere outside the primary vessel, but within the concrete cell, is nitrogen which also serves as a heat transfer medium to control the cell temperature. To provide space for the storage of test assemblies such as the OTA and CLIRA, 10 storage tubes (each approximately 43-1/4 ft long) are included near the center of the basket. This arrangement requires that the center of the primary vessel be quite deep. In this region, the primary vessel extends downward to elevation 501 ft 6 inches while the guard tank reaches 500 ft 4 inches. The floor of the cell is at 499 ft a inches which is 51 ft below the operating room floor. Storage positions are provided for 112 core components in the upper section of the storage basket. These positions are arranged in four circles, all of which are concentric with the test element array and the storage basket. The primary vessel and the guard tank are shaped to provide the necessary space with a minimum of excess volume. Both these vessels have a relatively small cylindrical lower section connected to a larger upper cylinder by a conical transition. The primary vessel is supported from a top flange by a vessel support structure. The guard tank is supported by a skirt which rests on a ledge at elevation 527 ft 2 inches. The skirt is an extension of the upper cylinder of the guard tank. The storage basket is supported by a gear-driven, mechanically indexed, ball bearing that rests on the bearing support, which in turn rests on the vessel support structure. The interior of the primary vessel above the sodium level is blanketed with argon at 6 inches of water gage pressure. The vessel is designed to allow the pressure to be inc