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1

Reactor pressure vessel nozzle  

DOE Patents (OSTI)

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

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

1994-10-04T23:59:59.000Z

2

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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...

3

Reactor pressure vessel vented head  

DOE Patents (OSTI)

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

Sawabe, James K. (San Jose, CA)

1994-01-11T23:59:59.000Z

4

Reactor pressure vessel vented head  

DOE Patents (OSTI)

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

Sawabe, J.K.

1994-01-11T23:59:59.000Z

5

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

6

Decommissioning: Reactor Pressure Vessel Internals Segmentation  

Science Conference Proceedings (OSTI)

Decommissioning a nuclear plant covers a wide variety of challenging projects. One of the most challenging areas is the removal and disposal of the reactor pressure vessel (RPV) and the RPV internals. This report describes commercial reactor pressure vessel segmentation projects that have been completed and discusses several projects that are still in the planning stages. The report also covers lessons learned from each project.

2001-10-11T23:59:59.000Z

7

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

8

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

9

Reactor pressure vessel with forged nozzles  

DOE Patents (OSTI)

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

Desai, Dilip R. (Fremont, CA)

1993-01-01T23:59:59.000Z

10

Nuclear reactor pressure vessel support system  

DOE Patents (OSTI)

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

Sepelak, George R. (McMurray, PA)

1978-01-01T23:59:59.000Z

11

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

12

Embrittlement of Nuclear Reactor Pressure Vessels  

Science Conference Proceedings (OSTI)

Boiler and Pressure Vessel Code Section III App. G Protection Against Nonductile Fracture (New York: American Society of Mechanical Engineers, 1986 ). 3.

13

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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,

14

Materials Reliability Program: Reactor Pressure Vessel Integrity Primer (MRP-278)  

Science Conference Proceedings (OSTI)

This primer is based on two earlier Electric Power Research Institute (EPRI) reports: Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity (TR-101975-T2) and Primer: Fracture Mechanics in the Nuclear Power Industry (NP-5792-SR, Rev. 1). The information in those earlier reports has been updated extensively and focuses on todays reactor pressure vessel (RPV) embrittlement, integrity, and plant license renewal issues. This RPV integrity ...

2010-06-09T23:59:59.000Z

15

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

16

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect

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

17

Materials Reliability Program: Reactor Pressure Vessel Integrity Training Module (MRP-286)  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern in ensuring continued safe operation. The shutdown of the Yankee Rowe plant, which occurred because of uncertainties related to embrittlement of the vessel, demonstrated the importance of adequately addressing embrittlement issues. Managing embrittlement involves the integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. Reactor vessel embrittlement management is an essential ...

2010-11-22T23:59:59.000Z

18

Qualification of In-Service Examination of the Yankee Rowe Reactor Pressure Vessel  

Science Conference Proceedings (OSTI)

An effective in-service examination of the reactor pressure vessel was an essential part of the restart program for the Yankee Atomic Power Company plant in Rowe, Massachusetts. This report describes development of an effective examination strategy, demonstration of performance of the examination procedures, and development of data on the distribution of flaws in reactor pressure vessels.

1993-01-01T23:59:59.000Z

19

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

Skalozubov, V I; Jarovoj, S S; Kochnyeva, V Yu

2012-01-01T23:59:59.000Z

20

Substantiation of Thermodynamic Criteria of Explosion Safety in Process of Severe Accidents in Pressure Vessel Reactors  

E-Print Network (OSTI)

The paper represents original development of thermodynamic criteria of occurrence conditions of steam-gas explosions in the process of severe accidents. The received results can be used for modelling of processes of severe accidents in pressure vessel reactors.

V. I. Skalozubov; V. N. Vashchenko; S. S. Jarovoj; V. Yu. Kochnyeva

2012-03-27T23:59:59.000Z

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

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

SciTech Connect

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

Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A. [Oak Ridge National Lab., TN (United States)

1993-08-01T23:59:59.000Z

22

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

Science Conference Proceedings (OSTI)

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

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

1995-03-01T23:59:59.000Z

23

Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels  

E-Print Network (OSTI)

in this paper. Keywords: Remote inspection, Service robot, Non-destructive test, Nuclear, Climbing robotWalking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor

Chen, Sheng

24

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

Science Conference Proceedings (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

25

Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel  

Science Conference Proceedings (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

26

Calvert Cliffs 1 Reactor Vessel: Pressurized Thermal Shock Analysis for a Small Steam Line Break  

Science Conference Proceedings (OSTI)

Analysis of this Maryland reactor revealed a wide safety margin in its two-loop Combustion Engineering PWR pressure vessel for transients caused by small steam line breaks. The study employed a new method for analyzing pressurized thermal shock effects that combines several EPRI computer codes.

1984-11-01T23:59:59.000Z

27

Ensuring the Performance of Nuclear Reactor Pressure Vessels for ...  

Science Conference Proceedings (OSTI)

The Light Water Reactor Sustainability Program is a collaborative program ... and in situ Mechanical Test Methods in the US Fusion Reactor Materials Program.

28

Conceptual Design of a Reactor Pressure Vessel and its Internals for a HPLWR  

Science Conference Proceedings (OSTI)

A design for the Reactor Pressure Vessel (RPV) and its internals for a HPLWR (High Performance Light Water Reactor) is presented. The RPV has been dimensioned using the pressure vessel code for nuclear power plants in Germany. In order to use conventional vessel materials such as 20 MnMoNi 5 5 (United States: SA 508), the vessel inner wall has to be kept only in contact with coolant at inlet temperature. Therefore, the hot coolant pipe connection from the steam plenum to the outlet is separated from the RPV inner wall using a thermal sleeve. The core inside the vessel rests on a support plate which is connected to the core barrel. The steam plenum is fixed on top of the core using support brackets which are attached to the adjustable steam outlet pipes. This way, the steam plenum rests on the outlet flanges of the lower vessel, while the core barrel is suspended at the closure head flange of the vessel to control thermal expansions between the internals and the RPV and to minimize thermal stresses. Both, inlet and outlet mass flows are separated via C-ring seals to prevent mixing. The control rod guides in the upper plenum are also suspended at the vessel flange and aligned inside the core barrel using centering pins. (authors)

Fischer, Kai [EnBW Kraftwerke AG, Kernkraftwerk Philippsburg, Rheinschanzinsel D-76661 Philippsburg (Germany); Starflinger, Joerg; Schulenberg, Thomas [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies P.O. Box 3640, D-76021 Karlsruhe (Germany)

2006-07-01T23:59:59.000Z

29

Fracture Toughness Characterization of Japanese Reactor Pressure Vessel Steels: Joint EPRI-CRIEPI RPV Embrittlement Studies  

Science Conference Proceedings (OSTI)

EPRI has examined five Japanese reactor pressure vessel steels to characterize the material properties over a complete temperature range, including the brittle/ductile transition region and the upper shelf typical of normal operation. The test results provide the unirradiated baseline needed for evaluating the effects of radiation embrittlement.

1993-07-01T23:59:59.000Z

30

Detection and characterization of indications in segments of reactor pressure vessels  

Science Conference Proceedings (OSTI)

Studies have been conducted to estimate flaw density in segments cut from light water reactor (LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (HSST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques and to compare the results with current assumptions related to probabilistic risk assessment. Both objectives were successfully completed. Ultrasonic techniques beyond those required by the 1986 edition of the ASME Boiler and Pressure Vessel Code were necessary for the detection and reporting of the detected discontinuities. Extra care and analysis must be exercised when conducting ultrasonic examination through cladding. The detection of the discontinuities in the arbitrarily selected sections implies that the Marshall report estimates (and others) are nonconservative for such small flaws. 8 refs., 9 figs.

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

1989-08-01T23:59:59.000Z

31

Materials Reliability Project: Benchmark Study of Reactor Pressure Vessel Integrity Probabilistic Computational Results Using the Fracture Analysis of Vessels – Oak Ridge (FAVOR) Software Code (MRP-371)  

Science Conference Proceedings (OSTI)

This report reports the results from the Fracture Analysis of Vessels – Oak Ridge (FAVOR) software analysis of three transients that simulated pressurized thermal shock events in pressurized water reactor (PWR) reactor pressure vessels (RPVs). It was determined that software modifications would be required to complete the probabilistic analyses for the wide range of flaw sizes and locations of interest in the study. Consequently, two software revisions were provided by EPRI to enable ...

2013-08-22T23:59:59.000Z

32

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

Science Conference Proceedings (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

33

Reactor Vessel Internals Inspection and Reactor Pressure Vessel Surveillance Program Summaries for R.E. Ginna and Nine Mile Point Unit 1  

Science Conference Proceedings (OSTI)

This report provides a summary of project activities involving the reactor pressure vessel and internals for the nuclear power plants included in the joint Electric Power Research Institute (EPRI), Department of Energy (DOE), and Constellation Energy Nuclear Group (CENG) Nuclear Plant Life Extension demonstration project.BackgroundThe project focused on continuing operations at two CENG nuclear units that are currently operating in their extended license ...

2013-12-18T23:59:59.000Z

34

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

Science Conference Proceedings (OSTI)

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

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

1987-01-01T23:59:59.000Z

35

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports  

SciTech Connect

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

36

Program on Technology Innovation: Weld Metals and Welding Processes for Fabrication of Advanced Light Water Reactor Pressure Vessels  

Science Conference Proceedings (OSTI)

Light water reactors have traditionally been constructed using roll-formed plates for the reactor pressure vessel (RPV) shells, which were assembled via horizontal and vertical seam welds. Weld filler metals often contained significant quantities of copper, other residual elements such as vanadium, and nonmetallic elements such as phosphorous and sulfur. Low-alloy steel weld filler metals of this chemical composition contributed to the degree of neutron radiation-induced embrittlement of vessel ...

2013-06-26T23:59:59.000Z

37

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

38

NP-3319, January 1984: Physically Based Regression Correlations of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs  

Science Conference Proceedings (OSTI)

The first physically based model for forecasting the embrittlement behavior of irradiation-damaged steels proved much more accurate than earlier models. Thisadvance offers utilities greater precision in establishing operating pressure-temperature limits for PWRs and in assessing the ability of reactor vessels to withstand pressurized thermal shock transients.BackgroundBombardment by high-energy neutrons in the belt line of nuclear reactors can ...

1984-01-31T23:59:59.000Z

39

Reactor pressure vessel integrity research at the Oak Ridge National Laboratory  

Science Conference Proceedings (OSTI)

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

Corwin, W.R.; Pennell, W.E.; Pace, J.V.

1995-12-31T23:59:59.000Z

40

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

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.
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41

TR-105696-R16 (BWRVIP-03) Revision 16: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines  

Science Conference Proceedings (OSTI)

This report provides the boiling water reactor (BWR) fleet with inspection options for all of the safety-related vessel internal components, and provides a stable mechanism for documenting the capability of the evolving inspection technology. It is the sole resource for internals inspection information for BWR ...

2013-12-14T23:59:59.000Z

42

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

Science Conference Proceedings (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

43

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

DOE Green Energy (OSTI)

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

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

2007-03-21T23:59:59.000Z

44

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

SciTech Connect

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

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01T23:59:59.000Z

45

Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity  

Science Conference Proceedings (OSTI)

For many reactor pressure vessels, embrittlement is the primary concern for continued safe operation. The shutdown of the Yankee Rowe plant because of uncertainties related to embrittlement of the vessel demonstrates the importance of adequately addressing embrittlement issues. Managing embrittlement requires integration, management, and implementation of diverse technical, regulatory, planning, and economic activities. An effective embrittlement management program will ensure vessel safety and reliabili...

1994-01-22T23:59:59.000Z

46

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

SciTech Connect

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

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

1992-03-01T23:59:59.000Z

47

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

Science Conference Proceedings (OSTI)

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

Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. (Oak Ridge National Lab., TN (United States))

1992-03-01T23:59:59.000Z

48

Study of the Neutron Flux and Dpa Attenuation in the Reactor Pressure-Vessel Wall  

Science Conference Proceedings (OSTI)

The study of the neutron flux and dpa attenuation in the reactor pressure vessel (PV) wall presented in this work was performed with state-of-the art methods currently used to determine PV fluxes, the BUGLE-96 cross-section library, and the iron displacement cross sections derived from ENDF/B-VI data. The calculations showed that the RG 1.99, Rev. 2, extrapolation formula predicts slower--and therefore conservative--attenuation of the neutron flux (E > 1MeV) in the PV wall. More importantly, the calculations gave slower attenuation of the dpa rate in the PV wall than the attenuation predicted by the formula. The slower dpa rate attenuation was observed for all the cases considered, which included two different PWRs, and several configurations obtained by varying the PV wall thickness and thermal shield thickness. For example, for a PV wall thickness of {approximately}24 cm, the calculated ratio of the dpa rate at 1/4 and 3/4 of the PV wall thickness to the dpa value on the inner PV surface is {approximately}14% and 19% higher, respectively, than predicted by the RG 1.99, Rev. 2, formula.

Remec, I.

1999-06-01T23:59:59.000Z

49

Materials Reliability Program: Destructive Examination of the North Anna 2 Reactor Pressure Vessel Head (MRP-198)  

Science Conference Proceedings (OSTI)

This document is the final of three reports concerning the nondestructive and destructive examinations of selected control rod drive mechanism (CRDM) penetrations from the decommissioned North Anna Unit 2 reactor vessel head (RVH). The phase-1 report of the EPRI-MRP (Materials Reliability Program) managed program described the selection and removal of penetrations from the decommissioned RVH and the penetration decontamination and laboratory nondestructive evaluation (NDE). The phase-2 report detailed th...

2006-11-13T23:59:59.000Z

50

Pressurized fluidized bed reactor  

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.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

51

Pressurized fluidized bed reactor  

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.

1996-03-19T23:59:59.000Z

52

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  

Science Conference Proceedings (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

53

Qualification of in-service examination of the Yankee Rowe reactor pressure vessel  

SciTech Connect

Technical support was provided to assist the Yankee Atomic Electric Company with their restart effort for the Yankee plant in Rowe, Massachusetts. Demonstration of adequate margin during a postulated pressurized thermal shock accident was an important part of the justification for restarting the plant, and effective inservice examination of the critical inner surface of the vessel in the beltline region was a key objective and a significant component of the safety analysis. This report discussed this inservice inspection.

Ammirato, F.; Kietzman, K.; Becker, L.; Ashwin, P.; Selby, G.; Krzywosz, K.; Findlan, S. (Electric Power Research Inst., Charlotte, NC (United States). Nondestructive Evaluation Center); Lance, J. (Yankee Atomic Electric Co., Bolton, MA (United States))

1992-12-01T23:59:59.000Z

54

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

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

55

Evolution of Design Methodologies for Next Generation of Reactor Pressure Vessels and Extensive Role of Thermal-Hydraulic Numerical Tools  

SciTech Connect

The thermal-hydraulic design of the first pressurized water reactors was mainly based on an experimental approach, with a large series of tests on the main equipment [control rod guide tubes, reactor pressure vessel (RPV) plenums, etc.] to check performance.Development of computational fluid dynamics codes and computers now allows for complex simulations of hydraulics phenomena. Provided adequate qualification, these numerical tools are an efficient means to determine hydraulics in the given design and to perform sensitivities for optimization of new designs. Experiments always play their role, first for qualification and then for validation at the last stage of the design. The design of the European Pressurized Water Reactor (EPR), jointly developed by Framatome ANP, Electricite de France (EDF), and the German utilities, is based on both hydraulics calculations and experiments handled in a complementary approach.This paper describes the collective effort launched by Framatome ANP and EDF on hydraulics calculations for the RPV of the EPR. It concerns three-dimensional calculations of RPV inlets, including the cold legs, the RPV downcomer and lower plenum, and the RPV upper plenum up to and including the hot legs. It covers normal operating conditions but also accidental conditions such as pressurized thermal shock in a small-break loss-of-coolant accident. Those hydraulics studies have provided much useful information for the mechanical design of RPV internals.

Bellet, Serge [Electricite de France - Septen (EDF) (France); Goreaud, Nicolas [Framatome ANP(France); Nicaise, Norbert [Framatome ANP (France)

2005-11-15T23:59:59.000Z

56

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

57

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

58

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

59

Results and Analyses of Irradiation/Anneal Experiments Conducted on Yankee Rowe Reactor Pressure Vessel Surrogate Materials: Yankee Atomic Electric Company Test Reactor Program  

Science Conference Proceedings (OSTI)

Many variables influence the response of reactor vessel steels to neutron irradiation. This study looks at the influence of irradiation temperature, steel heat treatment and microstructure, and nickel and phosphorus content on the irradiation response of high-copper reactor vessel steel. Also addressed are several studies evaluating the potential of thermal annealing to restore the mechanical properties of the steels tested.

1996-03-22T23:59:59.000Z

60

Synergetic Effect of Ni and Cu in French Reactor Pressure Vessel ...  

Science Conference Proceedings (OSTI)

Symposium, Materials and Fuels for the Current and Advanced Nuclear Reactors II ... A Rate-Theory Approach to Irradiation Damage Modeling with Random ...

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

Reactor vessel seal service fixture  

DOE Patents (OSTI)

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

Ritz, W.C.

1975-12-01T23:59:59.000Z

62

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

63

Nondestructive Evaluation: Procedure for Manual Phased Array Ultrasonic Examination of Reactor Pressure Vessel Nozzle-to-Shell Welds and Nozzle Inner Radius Regions  

Science Conference Proceedings (OSTI)

The nozzle inner radius and nozzle-to-shell welds in a reactor pressure vessel of nuclear power plants must be examined periodically using ultrasonic examination technology. Phased array ultrasonic technology has become available in a handheld, portable configuration. This technology could increase the speed of the examination and reduce radiation exposure. This phased array procedure is capable of supporting multiple phased array instruments and was originally qualified in 2008 using the OmniScan phased...

2010-10-13T23:59:59.000Z

64

Materials Reliability Program: Testing and Evaluation of Reactor Pressure Vessel Steel Plate Heat JRQ to Assess Through-Wall Attenua tion of Radiation Embrittlement (MRP-243)  

Science Conference Proceedings (OSTI)

The change in neutron energy spectrum through the wall of a reactor pressure vessel (RPV) requires the use of an exposure parameter or metric for assessing radiation embrittlement. This report looks at experimental fracture toughness and Charpy V-notch (CVN) data generated in a special International Atomic Energy Agency (IAEA) experiment designed to simulate an RPV wall of 190 mm thickness. These experimental data are compared with the current exposure metric of displacements per atom (dpa) coupled with ...

2008-12-23T23:59:59.000Z

65

Materials Reliability Program: Testing and Evaluation of Two Reactor Pressure Vessel Steels Irradiated to Assess Through-Wall Attenu ation of Radiation Embrittlement (MRP-203)  

Science Conference Proceedings (OSTI)

The change in neutron energy spectrum through the wall of a reactor pressure vessel (RPV) requires the use of an exposure parameter or metric for assessing radiation embrittlement. This report looks at experimental fracture toughness and Charpy V-notch data generated in a special International Atomic Energy Agency (IAEA) experiment designed to simulate an RPV wall of 180-mm thickness. These experimental data are compared with the current exposure metric of displacements per atom (dpa) coupled with an emb...

2006-10-04T23:59:59.000Z

66

Documentation of Probabilistic Fracture Mechanics Codes Used for Reactor Pressure Vessels Subjected to Pressurized Thermal Shock Loa ding: Parts 1 and 2  

Science Conference Proceedings (OSTI)

Pressurized thermal shock (PTS) can impact the safety and operability of PWR vessels with significant radiation embrittlement in the vessel walls. This report documents the results of probabilistic fracture mechanics analysis benchmark studies performed to validate the use of several codes for evaluating vessel PTS. Such benchmark studies provide the industry with a standard reference method for verifying probabilistic fracture mechanics codes used in PTS analyses.

1995-08-08T23:59:59.000Z

67

Cryogenic Pressure Vessels: Progress and Plans  

NLE Websites -- All DOE Office Websites (Extended Search)

Pressure Vessel workshop, LLNL, February 15, 2011, p. 1 Cryogenic Pressure Vessels: Progress and Plans Salvador Aceves, Gene Berry, Francisco Espinosa, Ibo Matthews, Guillaume...

68

Rancho Seco Reactor Vessel Segmentation Experience Report  

Science Conference Proceedings (OSTI)

This report documents the approach taken by Sacramento Municipal Utility District (SMUD) in the segmentation and disposal of the Reactor Vessel from the Rancho Seco Nuclear Generating Station (RSNGS). The location of the Rancho Seco plant placed major constraints on the shipping options available for large plant components (Steam Generators and Reactor Vessel). This report details the engineering evaluations leading to the segmentation and disposal of the Reactor Vessel (RV). It describes the key element...

2008-03-18T23:59:59.000Z

69

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

70

PRESSURE VESSEL FABRICATION USING T-1 STEEL  

SciTech Connect

The fabrication of pressure vessels using C-l steel is described. The welding, welding electrodes, explosionbulge test, and impact and fatigue properties for the pressure vessel are given. (W.L.H.)

Franco-Ferreira, E.A.

1957-11-14T23:59:59.000Z

71

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

Science Conference Proceedings (OSTI)

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

Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K. [Oak Ridge National Lab., TN (United States)

1998-03-01T23:59:59.000Z

72

Pressure vessel and piping codes  

SciTech Connect

Section III of the ASME Boiler and Pressure Vessel Code contains simplified design formulas for placing bounds on the plastic deformations in nuclear power plant piping systems. For Class 1 piping a simple equation is given in terms of primary load stress indices (B/sub 1/ and B/sub 2/) and nominal pressure and bending stresses. The B/sub 1/ and B/sub 2/ stress indices reflect the capacities of various piping products to carry load without gross plastic deformation. In this paper, the significance of the indices, nominal stresses, and limits given in the Code for Class 1 piping and corresponding requirements for Class 2 and Class 3 piping are discussed. Motivation behind recent (1978-1981) changes in the indices and in the associated stress limits is presented.

Moore, S.E.; Rodabaugh, E.C.

1982-11-01T23:59:59.000Z

73

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect

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

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

74

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

75

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

76

Experimental investigation of creep behavior of reactor vessel lower head  

SciTech Connect

The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling.

Chu, T.Y.; Pilch, M.; Bentz, J.H. [Sandia National Labs., Albuquerque, NM (United States); Behbahani, A. [NRC, Washington, DC (United States)

1998-03-01T23:59:59.000Z

77

Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) developed inspection and evaluation (I&E) guidelines for managing long-term aging reactor vessel internal components of pressurized water reactors (PWRs) reactor internals. Specifically, the guidelines are applicable to reactor vessel internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the reactor vessel.

2011-12-23T23:59:59.000Z

78

Materials Reliability Program: Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants (MR P-206)  

Science Conference Proceedings (OSTI)

This report presents inspection and evaluation guidelines for reactor vessel bottom-mounted nozzles in U.S. pressurized water reactor (PWR) plants.

2009-03-23T23:59:59.000Z

79

Materials Reliability Program: Phase II Work Plan for Developing a Risk-Informed Approach for Calculating Reactor Pressure Vessel He atup and Cooldown Operating Curves (MRP-195)  

Science Conference Proceedings (OSTI)

The current procedures for calculating pressure-temperature (P/T) limits for normal reactor heatup and cooldown are defined by the deterministic fracture mechanics methodology in Appendix G (in both Section XI and Section III) of the ASME Code. The recent pressurized thermal shock (PTS) reevaluation effort used a very thorough probabilistic fracture mechanics (PFM) evaluation to develop a technical basis to increase the PTS screening criteria. This same PFM methodology can be applied for evaluating norma...

2006-05-31T23:59:59.000Z

80

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

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

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

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

82

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

83

Hydrogen degradation and microstructural effects of the near-threshold fatigue resistance of pressure vessel steels  

E-Print Network (OSTI)

Safety of pressure vessels for applications such as coal conversion reactors requires understanding of the mechanism of environmentally-induced crack propagation and the mechanism by which process-induced microstructures ...

Fuquen-Molano, Rosendo

1982-01-01T23:59:59.000Z

84

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

NLE Websites -- All DOE Office Websites (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

85

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

Science Conference Proceedings (OSTI)

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

Vinson, Dennis

2010-06-01T23:59:59.000Z

86

Materials Reliability Program: San Onofre Nuclear Generating Station Reactor Vessel Internals Management Engineering Program (MRP-303)  

Science Conference Proceedings (OSTI)

All operating pressurized water reactors must have a reactor vessel internals aging management document in place by December 2011 according to the mandatory requirement under Nuclear Energy Institute (NEI) 03-08. This program should be developed to meet the guidance provided by Materials Reliability Program (MRP) -227, Rev. 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. For non-license renewal plants, the requirements are valid within the current license period, and the Elec...

2011-02-28T23:59:59.000Z

87

BWRVIP-167: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues. This BWR Vessel and Internals Project (BWRVIP) report provides BWR Issue Management Tables that identify, rank, and describe R&D gaps.

2007-03-20T23:59:59.000Z

88

BWRVIP-167NP, Revision 2: BWR Vessel and Internals Project, Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities face numerous ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables (IMTs) in the report are living documents that summarize the st...

2010-08-24T23:59:59.000Z

89

Vehicular Storage of Hydrogen in Insulated Pressure Vessels  

DOE Green Energy (OSTI)

This paper describes the development of an alternative technology for storing hydrogen fuel onboard automobiles. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept cryogenic liquid fuel, cryogenic compressed gas or compressed gas at ambient temperature. Insulated pressure vessels offer advantages over conventional H{sub 2} storage approaches. Insulated pressure vessels are more compact and require less carbon fiber than GH{sub 2} vessels. They have lower evaporative losses than LH{sub 2} tanks, and are much lighter than metal hydrides. After outlining the advantages of hydrogen fuel and insulated pressure vessels, the paper describes the experimental and analytical work conducted to verify that insulated pressure vessels can be used safely for vehicular H{sub 2} storage. The paper describes tests that have been conducted to evaluate the safety of insulated pressure vessels. Insulated pressure vessels have successfully completed a series of DOT, ISO and SAE certification tests. A draft procedure for insulated pressure vessel certification has been generated to assist in a future commercialization of this technology. An insulated pressure vessel has been installed in a hydrogen fueled truck and it is currently being subjected to extensive testing.

Aceves, S M; Berry, G D; Martinez-Frias, J; Espinosa-Loza, F

2005-01-03T23:59:59.000Z

90

BWRVIP-270, Revision 1: BWR Vessel and Internals Project, Compilation of Fluence Estimates for Boiling Water Reactor Materials  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) is an association of utilities focused on BWR vessel and internals issues. Many of the BWR internal components receive high exposure to neutron flux due to their proximity to the fuel in the Reactor Pressure Vessel (RPV). Identifying how predicted fluence values will impact the materials at these locations is a focus of the BWRVIP proactive materials strategy. As part of this approach, this report provides visual and tabular summaries ...

2013-12-09T23:59:59.000Z

91

International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings  

NLE Websites -- All DOE Office Websites (Extended Search)

challenges in harmonizing test protocols and requirements for compressed natural gas (CNG), hydrogen, and CNG-hydrogen (HCNG) blend pressure vessels and to define next steps for...

92

Damage analysis of composite pressure vessels using acoustic emission monitoring.  

E-Print Network (OSTI)

??Composite pressure vessels (CPVs) fabricated using a metal or plastic liner under a composite structural skin are commonly used for natural gas storage on road… (more)

Chou, H

2012-01-01T23:59:59.000Z

93

Reactor Vessel Head Disposal Campaign for Nuclear Management Company  

SciTech Connect

After establishing a goal to replace as many reactor vessel heads as possible - in the shortest time and at the lowest cost as possible - Nuclear Management Company (NMC) initiated an ambitious program to replace the heads on all six of its pressurized water reactors. Currently, four heads have been replaced; and four old heads have been disposed of. In 2002, NMC began fabricating the first of its replacement reactor vessel heads for the Kewaunee Nuclear Plant. During its fall 2004 refueling outage, Kewaunee's head was replaced and the old head was prepared for disposal. Kewaunee's disposal project included: - Down-ending, - Draining, - Decontamination, - Packaging, - Removal from containment, - On-Site handling, - Temporary storage, - Transportation, - Disposal. The next two replacements took place in the spring of 2005. Point Beach Nuclear Plant (PBNP) Unit 2 and Prairie Island Nuclear Generating Plant (PINGP) Unit 2 completed their head replacements during their scheduled refueling outages. Since these two outages were scheduled so close to each other, their removal and disposal posed some unique challenges. In addition, changes to the handling and disposal programs were made as a result of lessons learned from Kewaunee. A fourth head replacement took place during PBNP Unit 1's refueling outage during the fall of 2005. A number of additional changes took place. All of these changes and challenges are discussed in the paper. NMC's future schedule includes PINGP Unit 1's installation in Spring 2006 and Palisades' installation during 2007. NMC plans to dispose of these two remaining heads in a similar manner. This paper presents a summary of these activities, plus a discussion of lessons learned. (authors)

Hoelscher, H.L.; Closs, J.W. [Nuclear Management Company, LLC, 700 First Street, Hudson, WI 54016 (United States); Johnson, S.A. [Duratek, Inc., 140 Stoneridge Drive, Columbia, SC 29210 (United States)

2006-07-01T23:59:59.000Z

94

BWRVIP-241: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Ve ssel Shell Welds and Nozzle Blend Radii  

Science Conference Proceedings (OSTI)

This report documents supplemental analyses for boiling water reactor (BWR) reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii to address limitations imposed by the U.S. Nuclear Regulatory Commission (NRC) regarding the reduction of inspections specified in Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

2010-10-26T23:59:59.000Z

95

Reactor vessel using metal oxide ceramic membranes  

DOE Patents (OSTI)

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

Anderson, Marc A. (Madison, WI); Zeltner, Walter A. (Oregon, WI)

1992-08-11T23:59:59.000Z

96

Lightweight bladder lined pressure vessels - Energy Innovation ...  

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 ...

97

Materials Reliability Program, Reactor Vessel Head Boric Acid Corrosion Testing (MRP-165)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential...

2005-12-14T23:59:59.000Z

98

Corrosion Product Transport during Boiling Water Reactor and Pressurized Water Reactor Startups  

Science Conference Proceedings (OSTI)

Corrosion product transport to Pressurized Water Reactor (PWR) steam generators and to the Boiling Water Reactor (BWR) reactor vessel during startups is of increased interest due to reductions in feedwater transport rates during normal operation and the recent emphasis on minimizing total transport during the cycle. Reductions in transport will reduce deposition on the fuel and the tendency for hot spot formation in BWRs and reduce surface fouling and the tendency for formation of aggressive chemical sol...

2010-12-17T23:59:59.000Z

99

Materials Reliability Program: Evaluation of the Reactor Vessel Beltline Shell Forgings of Operating U.S. PWRs for Quasi-Laminar Indications (MRP-367)  

Science Conference Proceedings (OSTI)

In 2012, quasi-laminar indications were discovered in the beltline ring forgings of two Belgian pressurized water reactors (PWRs) during ultrasonic inspection (UT). This report assesses the implications of that discovery for U.S. reactor pressure vessels.BackgroundThe Doel 3 PWR has been operating in Belgium since 1982, Tihange 2 PWR since 1983. In 2012, UT of the ring forgings that constitute the cylindrical shells of the reactor pressure vessels (RPVs) ...

2013-11-14T23:59:59.000Z

100

Developing PWR Aging-Management Strategies for Reactor Vessel ...  

Science Conference Proceedings (OSTI)

AREVA Fuel Condition Index for a Pressurized Water Reactor .... Stress Corrosion Cracking Behavior near the Fusion Boundary of Dissimilar Weld Joint with ...

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

The Westinghouse Electric Corporation Reactor Vessel Radiation Surveillance Program  

Science Conference Proceedings (OSTI)

Westinghouse recognized that the disruption of the atomic lattice of metals by collision from energetic neutrons could alter the properties of the metals to such an extent that the changes could be of engineering significance. Furthermore, it was recognized that a physical-metallurgical phenomenon such as aging, both thermal and mechanical, also could alter the properties of a metal over its service life. Because of the potential changes in properties, reactor vessel radiation surveillance programs to monitor the effect of neutron radiation and other environmental factors on the reactor vessel materials during operational conditions over the life of the plant were initiated for Westinghouse plants with the insertion of reactor vessel material radiation surveillance capsules into the Yankee Atomic Company's Yankee Rowe plant in 1961.

Mayer, T.R.; Anderson, S.L.; Yanichko, S.E.

1983-01-01T23:59:59.000Z

102

Experiment Hazard Class 5.3 High Pressure Vessels  

NLE Websites -- All DOE Office Websites (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

103

Pressure vessel reliability as a function of allowable stress  

SciTech Connect

From Winter meeting of American Society of Mechanical Engineers; Detroit, Michigan, USA (11 Nov 1973). The probability of failure corresponding to specified levels of allowable design stress was calculated for pressure vessels designed in accordance with the ASME Boiler and Pressure Vessel Code. The analysis was performed for maximum shear stress failure and for cyclic stress failure. The significance of such failure prediction is ddscussed and a rationale for selecting an allowable stress is presented. Examples are presented that demonstrate the estimation of vessel failure probability as a function of load variation, strength variation, and design safety factor. (auth)

Arnold, H.G.

1972-01-01T23:59:59.000Z

104

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.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

105

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

106

Investigation of impulsively loaded pressure vessels  

SciTech Connect

Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

1963-10-15T23:59:59.000Z

107

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

SciTech Connect

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

108

A Flaw Tolerance Approach to Address Reactor Vessel Head Penetration Cracking Issue  

SciTech Connect

Nickel-based alloys and the associated welds are susceptible to Primary Water Stress Corrosion Cracking. In Pressurized Water Reactor nuclear power plants, the reactor vessel closure head upper penetration nozzles used for the Control Rod Drive Mechanisms and other instrumentation systems are made of such nickel-based alloys. Cracking and leakage have been observed in the upper head penetration nozzles in nuclear power plants worldwide. Such cracking and the resulting leakage is a degradation of the reactor vessel pressure boundary. Regulatory requirements have been issued by the Nuclear Regulatory Commission regarding periodic inspection of the susceptible areas to enable detection of indications and provide reasonable assurance of continued structural integrity for reactor vessel closure head. A flaw tolerance approach has been used in the disposition of detected indications to minimize outage delays, by performing up-front fracture mechanics evaluations for the common types of indications detected in the susceptible areas. Details of the flaw tolerance approach are presented in this paper. (authors)

Ng, C. K.; Jirawongkraisorn, S.; Swamy, S. [Westinghouse Electric Company, LLC, Nuclear Services Division, P. O. Box 158, Madison, PA 15663 (United States)

2006-07-01T23:59:59.000Z

109

Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum  

NLE Websites -- All DOE Office Websites (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

110

Analysis of the Catastrophic Rupture of a Pressure Vessel  

E-Print Network (OSTI)

occurred at a petroleum refinery in Chicago, killing 17 people and causing extensive property damage [1]. NBS was requested by the Occupational Safety and Health Administration (OSHA) to conduct an investigation into the failure of the pressure vessel that eyewitnesses identified as the initial source of the explosion and fire. This vessel was an amine absorber tower used to strip hydrogen sulfide from a process stream of propane and butane. The vessel was 18.8 m tall, 2.6 m in diameter, and constructed from 25 mm thick plates of type ASTM A516 Grade 70 steel. The investigation was complicated by the damage caused by the explosion and fire. The explosive force had been sufficient to propel the upper 14 m of the vessel a distance of 1 km from its original location,

unknown authors

1984-01-01T23:59:59.000Z

111

Fabrication of toroidal composite pressure vessels. Final report  

DOE Green Energy (OSTI)

A method for fabricating composite pressure vessels having toroidal geometry was evaluated. Eight units were fabricated using fibrous graphite material wrapped over a thin-walled aluminum liner. The material was wrapped using a machine designed for wrapping, the graphite material was impregnated with an epoxy resin that was subsequently thermally cured. The units were fabricated using various winding patterns. They were hydrostatically tested to determine their performance. The method of fabrication was demonstrated. However, the improvement in performance to weight ratio over that obtainable by an all metal vessel probably does not justify the extra cost of fabrication.

Dodge, W.G.; Escalona, A.

1996-11-24T23:59:59.000Z

112

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

113

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

114

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP  

DOE Patents (OSTI)

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

Puechl, K.H.

1963-09-24T23:59:59.000Z

115

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

116

Lightweight pressure vessels and unitized regenerative fuel cells  

DOE Green Energy (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

117

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

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

Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

1997-01-01T23:59:59.000Z

118

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

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

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

1997-12-16T23:59:59.000Z

119

Recent United States and International Experiences in Reactor Vessel and Internals Segmentation  

Science Conference Proceedings (OSTI)

The segmentation of reactor vessels and internals is one of the most challenging tasks in nuclear power plant decommissioning. Many experiences, lessons learned, and best practices have been gained through the execution of the first few reactor vessel and internals segmentation projects. The Electric Power Research Institute (EPRI) previously documented the experiences, lessons learned, best practices, and technologies used in decommissioning reactor vessel and internals segmentation projects in the Unit...

2011-11-29T23:59:59.000Z

120

PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS  

Science Conference Proceedings (OSTI)

Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

Hensel, S.

2012-03-27T23:59:59.000Z

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121

Material Reliability Program Technical Basis Document Concerning Irradiation-Induced Stress Relaxation and Void Swelling in Pressuri zed Water Reactor Vessel Internals Components (MRP-50)  

Science Conference Proceedings (OSTI)

Irradiation-induced swelling and irradiation-enhanced stress relaxation are two potential degradation mechanisms that could affect reactor vessel (RV) core internals components in pressurized water reactors (PWRs). This report describes current knowledge of these two potential degradation mechanisms, available relevant data and known functional relationships, and a qualitative assessment of these two mechanisms' combined and separate effects on PWR internals components.

2001-10-18T23:59:59.000Z

122

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

123

AN IBM 7090 FORTRAN PROGRAM FOR ASME UNFIRED PRESSURE VESSEL DESIGN AND PRELIMINARY COST ESTIMATION  

SciTech Connect

An IBM 7090 FORTRAN program was written for the preliminary design and cost estimation of unfired pressure vessels with or without a jacket. Both vessel and jacket designs conform to the 1959 ASME Boiler and Pressure Vessel Code, Section VIII, Unfired Pressure Vessels. Vessels and jackets from 5 in. pipe through 84 in. o.d. and 1/4 in. through 1 1/2 in. in metal thickness may be designed by this program as written. Total vessel cost is the sum of metal and fabrication costs, each on a weight basis. (auth)

Prince, C.E.; Milford, R.P.

1962-10-17T23:59:59.000Z

124

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables (MRP-205)  

Science Conference Proceedings (OSTI)

This report provides PWR Issue Management Tables (IMTs) that identify, prioritize, and describe R&D gaps related to degradation issues in PWR Reactor Pressure Vessels (RPVs), Reactor Internals, ASME Class 1 Piping Components, Pressurizers, and Steam Generators. An R&D "Gap" is identified whenever there are needs in the areas of degradation mechanism understanding, mitigation techniques, repair/replacement techniques, or inspection & evaluation technologies to provide reasonable assurance that a component...

2006-11-29T23:59:59.000Z

125

Role of ex-vessel interactions in determining the severe reactor-accident source term for fission products. [PWR; BWR  

SciTech Connect

The role fission-product release and aerosol generation outside the primary system can have in determining the severe reactor-accident source term is reviewed. Recent analytical and experimental studies of major causes of ex-vessel fission product release and aerosol generation are described. The ejection of molten-core debris from a pressurized-reactor vessel is shown to be a potentially large source of aerosols that has not been recognized in past severe-accident evaluations. A mechanistic model of fission-product release during core-debris interactions with concrete is discussed. Calculations with this model are compared to correlations of experimental data and previous estimates of ex-vessel fission-product release. Predictions with the mechanistic model agree quite well with the data correlations but do not agree at all well with estimates made in the past.

Powers, D.A.; Brockmann, J.E.; Bradley, D.R.; Tarbell, W.W.

1983-01-01T23:59:59.000Z

126

Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels  

SciTech Connect

In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K{sub Jc} values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K{sub Jc} data. By converting PCVN data to IT compact specimen equivalent K{sub Jc} data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K{sub Jc} database and the ASME lower bound K{sub Ic} curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K{sub Jc} with respect to K{sub Ic} in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K{sub Jc} data from PCVN specimens. 13 refs., 8 figs., 1 tab.

Sokolov, M.A.; Wallin, K.; McCabe, D.E.

1996-12-31T23:59:59.000Z

127

BWRVIP-44-A: BWR Vessel and Internals Project: Underwater Weld Repair of Nickel Alloy Reactor Vessel Internals  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This report describes work performed to qualify a flux-core welding process for use in repairing reactor internals at a water depth of up to 50 feet. A previous version of this report was published as BWRVIP-44 (EPRI report TR-108708). The current report, BWRVIP-44-A, incorporates changes proposed by the BWRVIP in response to U.S. Nuc...

2006-08-16T23:59:59.000Z

128

Analysis of Crack Development Involving a Pressure Vessel in a ...  

Science Conference Proceedings (OSTI)

The vessel is part of a by-product refining system comprising a synthetic natural gas production plant. The vessel processes a mixture of chemical species, ...

129

Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors  

SciTech Connect

This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

Love, E.F.; Pauley, K.A.; Reid, B.D.

1995-09-01T23:59:59.000Z

130

Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen and Natural Gas Storage  

Science Conference Proceedings (OSTI)

We are working on developing an alternative technology for storage of hydrogen or natural gas on light-duty vehicles. This technology has been titled insulated pressure vessels. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept either liquid fuel or ambient-temperature compressed fuel. Insulated pressure vessels offer the advantages of cryogenic liquid fuel tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for fuel liquefaction and reduced evaporative losses). The work described in this paper is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen or LNG. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining insulated pressure vessel certification.

Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F; Schaffer, R; Clapper, W

2002-05-22T23:59:59.000Z

131

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

Science Conference Proceedings (OSTI)

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

McCabe, D.E.

1999-09-01T23:59:59.000Z

132

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

133

Blowdown of hydrocarbons pressure vessel with partial phase separation  

E-Print Network (OSTI)

We propose a model for the simulation of the blowdown of vessels containing two-phase (gas-liquid) hydrocarbon fluids, considering non equilibrium between phases. Two phases may be present either already at the beginning of the blowdown process (for instance in gas-liquid separators) or as the liquid is formed from flashing of the vapor due to the cooling induced by pressure decrease. There is experimental evidence that the assumption of thermodynamic equilibrium is not appropriate, since the two phases show an independent temperature evolution. Thus, due to the greater heat transfer between the liquid phase with the wall, the wall in contact with the liquid experiences a stronger cooling than the wall in contact with the gas, during the blowdown. As a consequence, the vessel should be designed for a lower temperature than if it was supposed to contain vapor only. Our model is based on a compositional approach, and it takes into account internal heat and mass transfer processes, as well as heat transfer with ...

Speranza, Alessandro; 10.1142/9789812701817_0046

2011-01-01T23:59:59.000Z

134

PRESTRESSING A TWO-LAYER PRESSURE VESSEL BY CONTROLLED YIELDING OF THE INNER LAYER  

SciTech Connect

A method of designing a two-layer pressure vessel is presented wherein contact between the layers is produced by controlled yielding of the inner vessel by internal pressure. The amount of prestress depends upon the dimensions of the vessel, the properties of the material of construction, and the prestressing pressure. The method takes into account the actual stress-strain curve of the material and satisfies the rales of plastic flow with work hardening. (auth)

Schneider, R.W.

1964-04-01T23:59:59.000Z

135

PRESTRESSING A TWO-LAYER PRESSURE VESSEL BY CONTROLLED YIELDING OF THE INNER LAYER  

SciTech Connect

A method is presented for designing a two-layer pressure vessel wherein contact between the layers is produced by controlled yielding of the inner vessel by internal pressure. The amount of prestress depends upon the dimensions of the vessel, the properties of the material of construction, and the prestressing pressure. The method takes into account the actual stress-strain curve of the material and satisfies the rules of plastic flow with work hardening. (auth)

Schneider, R.W.

1964-01-29T23:59:59.000Z

136

Pressure vessel and piping codes. Technical basis for revised reference crack growth rate curves for pressure boundary steels in LWR environment  

SciTech Connect

Since the inception of the pressure vessel and piping codes the reference fatigue crack growth rate curves have been contained in Appendix A of Sect. XI. The curves have been designed to be applicable to carbon and low alloy pressure vessel steels exposed to either air or light water reactor coolant environments. Data obtained over the past several years have shown a different behavior of these steels in the light water reactor environment than that predicted by the present reference curve. A revised set of reference curves has been formulated, incorporating a new curve shape as well as a dependency of growth rate on R ratio (minimum load/maximum load). This work provides the background and justification for such a revision, details the methodology used to develop the revised curves, and includes an evalution of the adequacy and impact of the revised curves as compared with the single curve which they replace. 24 references.

Bamford, W.H.

1980-11-01T23:59:59.000Z

137

BWR Vessel and Internals Project: Quantitative Safety Assessment of BWR Reactor Internals (BWRVIP-09)  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June, 1994, is an association of utilities focused exclusively on BWR vessel and internals issues. This BWRVIP report documents the results of a quantitative safety assessment conducted to evaluate the safety significance of failures of certain BWR internal components.

1997-02-20T23:59:59.000Z

138

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

139

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

Science Conference Proceedings (OSTI)

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

140

Helium leak testing of a radioactive contaminated vessel under high pressure in a contaminated environment  

SciTech Connect

At ANL-W, with the shutdown of EBR-II, R&D has evolved from advanced reactor design to the safe handling, processing, packaging, and transporting spent nuclear fuel and nuclear waste. New methods of processing spent fuel rods and transforming contaminated material into acceptable waste forms are now in development. Storage of nuclear waste is a high interest item. ANL-W is participating in research of safe storage of nuclear waste, with the WIPP (Waste Isolation Pilot Plant) site in New Mexico the repository. The vessel under test simulates gas generated by contaminated materials stored underground at the WIPP site. The test vessel is 90% filled with a mixture of contaminated material and salt brine (from WIPP site) and pressurized with N2-1% He at 2500 psia. Test acceptance criteria is leakage < 10{sup -7} cc/seconds at 2500 psia. The bell jar method is used to determine leakage rate using a mass spectrometer leak detector (MSLD). The efficient MSLD and an Al bell jar replaced a costly, time consuming pressure decay test setup. Misinterpretation of test criterion data caused lengthy delays, resulting in the development of a unique procedure. Reevaluation of the initial intent of the test criteria resulted in leak tolerances being corrected and test efficiency improved.

Winter, M.E.

1996-10-01T23:59:59.000Z

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.


141

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

SciTech Connect

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

Dunn, F.E.

1987-01-01T23:59:59.000Z

142

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

E-Print Network (OSTI)

RISÃ?-M- 2186 INTERPRETATIOM OF STRAIN HBASUREMEMTS ON NUCLEAR PRESSURE VESSELS Svend Ib Andersen Preben Engbzk Abstract. Selected results from strain measurements on 4 nuclear pressure vessels procedure before and after the test as well as a detailed knowledge of the behaviour of the signal from

143

Repair Technology for Degraded Pressure Vessel and Heat Exchanger Shells: RRAC Task 91  

Science Conference Proceedings (OSTI)

The ability to repair pressure vessels and heat exchangers offers utilities significant cost savings compared to replacing these components. This is especially the case if outage time and loss of production are factored into the cost of replacement. This guide provides a review of various current and proposed repair methods that can be used for pressure vessel and heat exchanger applications.

2002-12-02T23:59:59.000Z

144

Computation of initial stage of RBMK reactor fuel channel vessel rupture  

SciTech Connect

Objective of this work is estimation of temperature and time characteristics for rupture of the zirconium pipe which is the RBMK reactor fuel channel (FC) vessel under emergencies. As an emergency the zirconium pipe temperature rise process is considered which results in loss of pipe material strength properties and pipe rupture under the action of internal pressure P=80MPa. The work was carried out under Task Order 007 of University of California - VNIIEF Subcontract No. 0002P0004-95. The problem formulation is stated in Protocol (Task 3, Appendix 3) of the Russian-American Workshop which was held in December, 1994 in Los Alamos. Physical-mechanical and geometry characteristics of structure elements (FC vessel with graphite ring and graphite slug) are presented by NIKIET. The temperature mode of the structure is taken in conformity with the NIKIET data obtained with the RELAP5/MOD3 code. Numerical simulation of structure element behavior in an emergency is performed using the DRAKON program comlex oriented to solving strength problems for complex spatial structures at intense dynamic loading. The {open_quotes}DRAKON{close_quotes} program complex is described and compared with similar western codes in its capabilities.

Pevnitsky, A.V.; Solovyev, V.P.; Abakumov, A.I. [and others

1995-12-31T23:59:59.000Z

145

CONVECTION REACTOR  

DOE Patents (OSTI)

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

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

1960-03-22T23:59:59.000Z

146

Comparison of Pressure Vessel Design and Inspection Requirements as Defined by ASME Code and Germany's TRD Code  

Science Conference Proceedings (OSTI)

This report compares the American Society of Mechanical Engineers (ASME) Code with the German TRD Code for pressure vessel engineering, fabrication, inspection, and other pressure vessel processes. The report compares calculations of minimum required wall thickness for pressure vessels such as boiler tubes, pipes, headers, and drums. It also compares material allowable stress values and reviews the major materials permitted by both codes for use in pressure vessel engineering and manufacturing. The repor...

1994-09-22T23:59:59.000Z

147

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

Hunsbedt, A.; Boardman, C.E.

1995-04-11T23:59:59.000Z

148

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

SciTech Connect

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

149

Materials Reliability Program: Qualification Protocol for Pressurized Water Reactor Upper Head Penetration Ultrasonic Examinations-- -2010 Update (MRP-234)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) has directed the Inspection Issues Task Group (ITG) to establish a qualification program for the examination of pressurized water reactor (PWR) reactor pressure vessel upper head penetrations. This new qualification program is being implemented to provide the utilities with a consistent and reliable examination approach for the upper head penetrations. The program will provide assurance that flaws of similar size and location will be detected reliably throughout th...

2010-07-30T23:59:59.000Z

150

Materials Reliability Program: Qualification Protocol for Pressurized Water Reactor Upper Head Penetration Ultrasonic Examinations - - 2012 Update (MRP-311, Revision 1)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) has directed the Inspection Technical Advisory Committee to establish a qualification program for the examination of pressurized water reactor (PWR) reactor pressure vessel upper head (RPVUH) penetrations. This qualification program is being implemented to provide the utilities with a consistent and reliable examination approach for upper head penetrations. The program will provide assurance that flaws of similar size and location will be detected reliably ...

2012-11-28T23:59:59.000Z

151

Materials Reliability Program: Qualification Protocol for Pressurized Water Reactor Upper Head Penetration Ultrasonic Examinations 2 011 Update (MRP-311)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) has directed the Inspection Technical Advisory Committee (TAC) to establish a qualification program for the examination of pressurized water reactor (PWR) reactor pressure vessel upper head penetrations (RPVUHs). This new qualification program was implemented to provide the utilities with a consistent and reliable examination approach for upper head penetrations. The program will provide assurance that flaws of similar size and location will be detected reliably th...

2011-09-28T23:59:59.000Z

152

DEVELOPMENT OF ASME SECTION X CODE RULES FOR HIGH PRESSURE COMPOSITE HYDROGEN PRESSURE VESSELS WITH NON-LOAD SHARING LINERS  

DOE Green Energy (OSTI)

The Boiler and Pressure Vessel Project Team on Hydrogen Tanks was formed in 2004 to develop Code rules to address the various needs that had been identified for the design and construction of up to 15000 psi hydrogen storage vessel. One of these needs was the development of Code rules for high pressure composite vessels with non-load sharing liners for stationary applications. In 2009, ASME approved new Appendix 8, for Section X Code which contains the rules for these vessels. These vessels are designated as Class III vessels with design pressure ranging from 20.7 MPa (3,000 ps)i to 103.4 MPa (15,000 psi) and maximum allowable outside liner diameter of 2.54 m (100 inches). The maximum design life of these vessels is limited to 20 years. Design, fabrication, and examination requirements have been specified, included Acoustic Emission testing at time of manufacture. The Code rules include the design qualification testing of prototype vessels. Qualification includes proof, expansion, burst, cyclic fatigue, creep, flaw, permeability, torque, penetration, and environmental testing.

Rawls, G.; Newhouse, N.; Rana, M.; Shelley, B.; Gorman, M.

2010-04-13T23:59:59.000Z

153

Guidelines for Managing Reactor Vessel Material Uncertainties: Part 1: General Approach Part 2: Implementation Guide  

Science Conference Proceedings (OSTI)

Uncertainties about reactor vessel material toughness properties can be a concern for utilities when characterizing vessel integrity. In addition, recent emphasis on variability in material chemistry and initial toughness properties has added to regulatory concerns. This two-part guidelines document provides a general approach (Part 1) for dealing with weld metal property variability and material uncertainties and demonstrates examples of different approaches (Part 2) for dealing with these uncertainties...

1997-04-30T23:59:59.000Z

154

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

155

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

Science Conference Proceedings (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 ({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

156

Pipeline and Pressure Vessel R&D under the Hydrogen Regional...  

NLE Websites -- All DOE Office Websites (Extended Search)

Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania Kevin L. Klug, Ph.D. 25 September 2007 DOE Hydrogen Pipeline Working Group...

157

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests  

DOE Green Energy (OSTI)

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

1991-08-01T23:59:59.000Z

158

Acoustic emission: flaw relationship for inservice monitoring of nuclear reactor pressure boundaries. [PWR; BWR  

Science Conference Proceedings (OSTI)

The objective of the acoustic emission (AE)/flaw characterization program is to provide an experimental feasibility evaluation of using the AE method on a continuous basis (during operation and during hydrotest) to detect and analyze flaw growth in reactor pressure vessels and primary piping. This effort is based on the philosophy that AE shows demonstrated capability for being a valuable addition to current nondestructive inspection (NDI) methods with unique capability for continuous monitoring, high sensitivity and remote flaw location.

Not Available

1981-10-01T23:59:59.000Z

159

White Paper on Reactor Vessel Integrity Requirements for Level A and B Conditions  

Science Conference Proceedings (OSTI)

The ASME Section XI Task Group recommends that the Reactor Vessel Integrity Requirements of the ASME code be updated in several areas to reflect current technology. This report reviews current regulations to identify areas of conservatism and potential nonconservatism and to provide recommendations for future code improvements.

1993-03-09T23:59:59.000Z

160

Materials Degradation Issues in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

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

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

162

Pressurized Water Reactor Secondary Water Chemistry Guidelines – Revision 6  

Science Conference Proceedings (OSTI)

State-of-the-art water chemistry programs reduce equipment corrosion and enhance steam generator reliability. A committee of industry experts prepared these revised "Pressurized Water Reactor Secondary Water Chemistry Guidelines" to incorporate the latest field and laboratory data on secondary system corrosion and performance issues. Pressurized water reactor (PWR) operators can use these guidelines to update their secondary water chemistry programs.

2004-12-13T23:59:59.000Z

163

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

SciTech Connect

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

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

1984-04-01T23:59:59.000Z

164

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

SciTech Connect

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

165

Metallographic and hardness examinations of TMI-2 lower pressure vessel head samples  

SciTech Connect

Fifteen steel samples were removed from the lower pressure vessel head of the damaged TMI-2 nuclear reactor to assess the thermal threat to the head posed by 15 to 20 metric tons of molten core debris relocating there during the accident. Full sections of thirteen of the samples and partial sections of the other two samples underwent hardness and metallographic examinations at the Idaho National Engineering Laboratory. These examinations have shown that eleven of the fifteen samples did not exceed the ferrite-austenite transformation temperature of 727 C during the accident. The remaining four samples did show evidence of having a much more severe thermal history. The samples from core grid positions F-10 and G-8 are believed to have experienced temperatures of 1,040 to 1,060 C for about 30 minutes. Samples from positions E-8 and E-6 appear to have been subjected to 1,075 to 1,100 C for approximately 30 minutes.

Korth, G. E. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-03-01T23:59:59.000Z

166

Welding and Repair Technology Center: Repair Technology for Degraded Pressure Vessel and Heat Exchanger Shells  

Science Conference Proceedings (OSTI)

BackgroundPressure vessels and heat exchangers are subject to a number of degradation mechanisms that can cause thinning of component walls and deterioration of internal components. With many repair options available, the Electric Power Research Institute (EPRI) Welding and Repair Technology Center (WRTC) has developed this report to assist operations and engineering personnel who are faced with defective or failed vessel components. Many available repair options allow ...

2013-10-30T23:59:59.000Z

167

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

168

Materials Reliability Program: An Assessment of the Control Rod Drive Mechanism (CRDM) Alloy 600 Reactor Vessel Head Penetration PWS CC Remedial Techniques (MRP-61)  

Science Conference Proceedings (OSTI)

Service experience over the past decade with control rod drive mechanism (CRDM) penetrations in pressurized water reactors (PWRs) worldwide confirmed primary water stress corrosion cracking (PWSCC) in alloy 600 base metal at several plants. This report summarizes the evaluations and results of an autoclave-accelerated stress corrosion cracking (SCC) test program designed to assess the effectiveness of selected surface remedial techniques to mitigate alloy 600 PWSCC in PWR vessel head penetration base and...

2003-07-14T23:59:59.000Z

169

Failure analysis of ETAC (Enrichment Technology Applications Center) pressure vessel  

Science Conference Proceedings (OSTI)

This report presents the results of an investigation into the failure of a graphite-epoxy composite cylinder. It investigates the quality of the as-fabricated cylinder and provides a verification of compressive material property input used in its design. The design is reevaluated in terms of the adjusted composition and material property input for its suitability for 18,000-psi pressure applications. A comparison between the composition and layup of a cylinder manufactured by Hitco is also provided, as well as the results of a pressurization test of an identical ETAC cylinder tested by the Naval Ocean Systems Center.

Frame, B.J.

1987-06-01T23:59:59.000Z

170

Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures  

SciTech Connect

Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X.

Swindeman, R.W.; Brinkman, C.R.

1981-01-01T23:59:59.000Z

171

Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors- Request for Relief"  

E-Print Network (OSTI)

2) NRC letter to TVA dated January 27, 2006, 'Watts Bar Nuclear Plant, Unit 1- Request for Relaxation from the First Revised NRC Order EA-03-009, Dated February 20, 2004, Deferral of Non-visual Nondestructive Examinations (TAC No. MC8543)" 3) TVA letter to NRC dated March 3, 2008, 'Watts Bar Nuclear Plant (WBN)

Mike Skaggs; U. S. Nuclear; Regulatory Commission; Tennessee Valley Authority

2008-01-01T23:59:59.000Z

172

BWRVIP-239: BWR Vessel and Internals Project, Updated Evaluation of the Integrated Surveillance Program (ISP) Capsule Withdrawal Sch edule  

Science Conference Proceedings (OSTI)

This report evaluates updated reactor pressure vessel and surveillance capsule fluence data for potential impacts on the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP) capsule withdrawal schedule.

2010-07-16T23:59:59.000Z

173

AUTHORIZATION OF THE TROJAN REACTOR VESSEL PACKAGE FOR ONE-TIME SHIPMENT FOR DISPOSAL (1)  

E-Print Network (OSTI)

Request Commission approval, by negative consent, for the staff to grant two specific exemptions from package test requirements specified in 10 CFR Part 71 for the Trojan Reactor Vessel Package (TRVP), and to authorize the TRVP for one-time transport for disposal. BACKGROUND: Portland General Electric Company (PGE) has requested approval of the TRVP (including internals) for transport to the disposal facility operated by US

L. Joseph; Callan Executive; Director Operations; John R. Cook

1998-01-01T23:59:59.000Z

174

Materials Reliability Program: Reactor Vessel Head Boric Acid Corrosion Testing (MRP-199)  

Science Conference Proceedings (OSTI)

PWR coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential stages of boric acid corros...

2007-06-27T23:59:59.000Z

175

Materials Reliability Program: Utility Preparation for Nondestructive Evaluation of Reactor Vessel Upper Head Penetrations (MRP-360)  

Science Conference Proceedings (OSTI)

The purpose of this report is to provide nuclear power plant owners with recommendations for planning and executing reactor vessel upper head (RVUH) penetration examinations in a manner that will minimize the occurrence of human errors while maximizing the probability of success. RVUH penetrations include control rod drive mechanism, control element drive mechanism, in-core instrumentation, and vent line penetrations. These encompass the standard nomenclatures for domestic utilities but might not ...

2013-12-19T23:59:59.000Z

176

Trojan Nuclear Power Plant Reactor Vessel and Internals Removal: Trojan Nuclear Plant Decommissioning Experience  

Science Conference Proceedings (OSTI)

One goal of the EPRI Decommissioning Technology Program is to capture the growing utility experience in nuclear plant decommissioning activities for the benefit of other utilities facing similar challenges in the future. This report provides historical information on the background, scope, organization, schedule, cost, contracts, and support activities associated with the Trojan Nuclear Plant Reactor Vessel and Internals Removal (RVAIR) Project. Also discussed are problems, successes, and lessons learned...

2000-10-16T23:59:59.000Z

177

Reactor core isolation cooling system  

DOE Patents (OSTI)

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Cooke, F.E.

1992-12-08T23:59:59.000Z

178

Reactor core isolation cooling system  

DOE Patents (OSTI)

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

179

Test of 6-inch-thick pressure vessels. Series 1: intermediate test vessels V-1 and V-2  

SciTech Connect

The intermediate vessel tests have been subdivided into four seriesi flaws in cylindrical vessels, A508, class 2 forging steel-two vessels; flaws in cylindrical vessels with longitudinal weld seams, A508, class 2 forging steel, submerged-arc welds-three vessels; flaws in cylindrical vessels wlth longitudinal weld seams, A533, grade B, class l plate steel, submerged-arc weld-two vessels; and cylindrical vessels with radially attached nozzles, vessels of A508, chass 2 forging steel and A533, grade B, class 1 plate steel; nozzle of A508 class 2 forging steel-three vessels. A comprehensive description of the pertinent factors considered in the design of the vessels is presented. Construction of the test facility and documentation of test results and fracture predictions are included. Emphasis is placed on providing the test results in such a manner that they form a resource for amy investigators interested in the problem of fracture. (auth)

Derby, R.W.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.; Witt, F.J.

1974-02-01T23:59:59.000Z

180

Test Results Using a Bell Jar to Measure Containment Vessel Pressurization  

SciTech Connect

A bell jar is used to determine containment vessel pressurization due to outgassing of plutonium materials. Fifteen food cans containing plutonium bearing materials, including plutonium packaged in direct contact with plastic and plutonium contaminated enriched oxide have been tested to date.

Hensel, S.J.

2002-05-10T23:59:59.000Z

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

Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure  

E-Print Network (OSTI)

Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania Kevin L. Klug, Ph.D. 25 September 2007 DOE Hydrogen Pipeline Working Group Meeting, Aiken, SCPerComp Engineering Inc. (HEI) ­ American Society Of Mechanical Engineers (ASME) ­ Pipeline Working Group (PWG) #12

182

Reactor technology. Progress report, January-March 1980  

Science Conference Proceedings (OSTI)

Progress is reported concerning space reactor (SPAR) electric power supply; GCFR reactor safety experiments; structural analysis of HTGR, PWR, and BWR containment vessels and pressure vessels; heat pipe technology development; and nuclear criticality experiments and safety.

Breslow, M.; Sullivan, S. (eds.)

1980-06-01T23:59:59.000Z

183

Actinide minimization using pressurized water reactors  

E-Print Network (OSTI)

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01T23:59:59.000Z

184

Some mechanistic observations on the crack growth characteristics of pressure vessel and piping steels in PWR environment  

SciTech Connect

The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288/sup 0/C (550/sup 0/F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.

Bamford, W.H.; Moon, D.M.

1979-01-01T23:59:59.000Z

185

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents (OSTI)

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

Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

1992-01-01T23:59:59.000Z

186

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

187

Pressure vessel sliding support unit and system using the sliding support unit  

DOE Patents (OSTI)

Provided is a sliding support and a system using the sliding support unit. The sliding support unit may include a fulcrum capture configured to attach to a support flange, a fulcrum support configured to attach to the fulcrum capture, and a baseplate block configured to support the fulcrum support. The system using the sliding support unit may include a pressure vessel, a pedestal bracket, and a plurality of sliding support units.

Breach, Michael R.; Keck, David J.; Deaver, Gerald A.

2013-01-15T23:59:59.000Z

188

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)

189

BWRVIP-189: BWR Vessel and Internals Project, Evaluation of RAMA Fluence Methodology Calculational Uncertainty  

Science Conference Proceedings (OSTI)

This report documents the overall calculational uncertainty associated with the application of the Radiation Application Modeling Application (RAMA) Fluence Methodology to BWR reactor pressure vessel fluence evaluations.

2008-07-07T23:59:59.000Z

190

Overcoming Solubility Limitations to Zinc Addition in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Zinc addition to the reactor coolant system (RCS) of a pressurized water reactor (PWR) is being used for dose rate reduction and primary water stress corrosion cracking (PWSCC) mitigation. This report summarizes results of aqueous zinc oxide solubility experiments from 150 to 350 degrees Celsius (302 to 662 degrees Fahrenheit). These experiments were performed to develop quantitative models of solubility and aqueous-phase solute speciation behavior as functions of temperature, pH, and solution compositio...

2001-11-29T23:59:59.000Z

191

Measurement of Wind Waves and Wave-Coherent Air Pressures on the Open Sea from a Moving SWATH Vessel  

Science Conference Proceedings (OSTI)

The design and implementation on a Small Waterline Area Twin Hull (SWATH) vessel of a complete system for measuring the directional distribution of wind waves and the concomitant fluctuations of air pressure and wind speed immediately above them ...

Mark A. Donelan; Fred W. Dobson; Hans C. Graber; Niels Madsen; Cyril McCormick

2005-07-01T23:59:59.000Z

192

Weld Repair of a Stamped Pressure Vessel in a Radiologically Controlled Zone  

Science Conference Proceedings (OSTI)

In September 2012 an ASME B&PVC Section VIII stamped pressure vessel located at the DOE Hanford Site Effluent Treatment Facility (ETF) developed a through-wall leak. The vessel, a steam/brine heat exchanger, operated in a radiologically controlled zone (by the CH2MHill PRC or CHPRC), had been in service for approximately 17 years. The heat exchanger is part of a single train evaporator process and its failure caused the entire system to be shut down, significantly impacting facility operations. This paper describes the activities associated with failure characterization, technical decision making/planning for repair by welding, logistical challenges associated with performing work in a radiologically controlled zone, performing the repair, and administrative considerations related to ASME code requirements.

Cannell, Gary L. [Fluor Enterprises, Inc.; Huth, Ralph J. [CH2MHill Plateau Remediation Company; Hallum, Randall T. [Fluor Government Group

2013-08-26T23:59:59.000Z

193

INSTRUMENT TRANSMITTERS FOR HIGH-PRESSURE, AQUEOUS, NUCLEAR REACTORS  

SciTech Connect

A review of the criteria involved in the selection of primary sensing elements for the measurement of process variables in high-pressure, aqueous, nuclear reactors is presented. Some acceptable types of sensing elements now in use at ORNL are described. (auth)

Moore, R.L.

1958-10-28T23:59:59.000Z

194

Pressurized Water Reactor Secondary Water Chemistry Guidelines - Revision 7  

Science Conference Proceedings (OSTI)

State-of-the-art water chemistry programs reduce equipment corrosion and enhance steam generator reliability. A committee of industry experts prepared these revised PWR Secondary Water Chemistry Guidelines to incorporate the latest field and laboratory data on secondary system corrosion and performance issues. Pressurized water reactor (PWR) operators can use these guidelines to update their secondary water chemistry programs.

2009-02-17T23:59:59.000Z

195

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

DOE Green Energy (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

196

Sampling Considerations for Monitoring Corrosion Products in the Reactor Coolant System in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Chemistry sampling of the reactor coolant system (RCS) of pressurized water reactors (PWRs) can provide significant information regarding the health of the primary system. Timely detection of increased corrosion product concentrations will aid in evaluating any risks associated with the onset of an axial offset anomaly, increased shutdown releases, increased out-of-core dose rates, or increased personnel doses. This report provides recommendations for improved RCS sampling.

2006-06-19T23:59:59.000Z

197

Condensate Polishing Guidelines for Pressurized Water Reactor and Boiling Water Reactor Plants - 2004 Revision  

Science Conference Proceedings (OSTI)

Successful condensate polishing allows more reliable operation of nuclear units by maintaining control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generators and the boiling water reactor (BWR) and recirculation system. This report presents revisions of EPRI's 1997 nuclear industry consensus guidelines for the design and operation of deep bed and filter demineralizer condensate polishers. These guidelines are consistent with the 2000 revisions of EPRI's "BWR W...

2004-03-16T23:59:59.000Z

198

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS  

DOE Patents (OSTI)

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16T23:59:59.000Z

199

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.

200

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.


201

Upper internals arrangement for a pressurized water reactor  

DOE Patents (OSTI)

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

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

2013-07-09T23:59:59.000Z

202

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Christy, R.F.

1958-07-15T23:59:59.000Z

203

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II  

Science Conference Proceedings (OSTI)

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Wang, C.Y.

1993-06-01T23:59:59.000Z

204

BLENDED CALCIUM ALUMINATE-CALCIUM SULFATE CEMENT-BASED GROUT FOR P-REACTOR VESSEL IN-SITU DECOMMISSIONING  

SciTech Connect

The objective of this report is to document laboratory testing of blended calcium aluminate - calcium hemihydrate grouts for P-Reactor vessel in-situ decommissioning. Blended calcium aluminate - calcium hemihydrate 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 which has a pH greater than 12.4. In addition, blended calcium aluminate - calcium hemihydrate cement compositions can be formulated such that the primary cementitious phase is a stable crystalline material. A less alkaline material (pH {<=} 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 [Wiersma, 2009a and b, Wiersma, 2010, and Serrato and Langton, 2010]. Information concerning access points into the P-Reactor vessel and amount of aluminum metal in the vessel is provided elsewhere [Griffin, 2010, Stefanko, 2009 and Wiersma, 2009 and 2010, Bobbitt, 2010, respectively]. Radiolysis calculations are also provided in a separate document [Reyes-Jimenez, 2010].

Langton, C.; Stefanko, D.

2011-03-10T23:59:59.000Z

205

Process and apparatus for adding and removing particles from pressurized reactors  

DOE Patents (OSTI)

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

Milligan, John D. (Little Silver, NJ)

1983-01-01T23:59:59.000Z

206

FOOD IRRADIATION REACTOR  

DOE Patents (OSTI)

An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

Leyse, C.F.; Putnam, G.E.

1961-05-01T23:59:59.000Z

207

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

SciTech Connect

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

208

Investigation of leaks in fiberglass-reinforced pressure vessels by direct observation of hollow fibers in glass cloth  

SciTech Connect

A simple method of visual observation of hollow fibers within fiberglass cloth has been developed. This visualization can aid in determining the contribution these fibers make toward leaks observed in fiberglass-reinforced epoxy resin pressure or vacuum vessels. Photographs and frequency data of these hollow fibers are provided. 3 figs.

McAdams, J.

1988-01-01T23:59:59.000Z

209

End-of-Life Rod Internal Pressures in Spent Pressurized Water Reactor Fuel  

Science Conference Proceedings (OSTI)

The end-of-reactor-life (EOL) rod internal pressure (RIP) is the primary protagonist for several evolutionary changes during long-term dry storage, which affect cladding resistance to failure when spent fuel assemblies are subjected to normal and accident conditions of transport. At the maximum temperature attained, either during vacuum drying or dry storage, EOL RIP determines the maximum stress state in the fuel rod cladding, which in turn sets the initial conditions for potential time-dependent ...

2013-12-17T23:59:59.000Z

210

Using SA508/533 for the HTGR Vessel Material  

SciTech Connect

This paper examines the influence of High Temperature Gas-cooled Reactor (HTGR) module power rating and normal operating temperatures on the use of SA508/533 material for the HTGR vessel system with emphasis on the calculated times at elevated temperatures approaching or exceeding ASME Code Service Limits (Levels B&C) to which the reactor pressure vessel could be exposed during postulated pressurized and depressurized conduction cooldown events over its design lifetime.

Larry Demick

2012-06-01T23:59:59.000Z

211

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect

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

212

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

SciTech Connect

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

213

Transmutation rates in the annulus gas of pressure tube water reactors.  

E-Print Network (OSTI)

??CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an… (more)

Ahmad, Mohammad Mateen

2011-01-01T23:59:59.000Z

214

NRC proposes new reactor embrittlement regulations  

SciTech Connect

The Nuclear Regulatory Commission has proposed amending its regulations regarding embrittlement of pressure vessels in commercial nuclear power reactors. The amendments, which include clarifying the pressurized thermal shock requirements, changing the fracture toughness requirements and establishing new requirements for thermal annealing of a reactor pressure vessel, were prompted in part by NRC`s 1990 inspection of the reactor pressure vessel at the Yankee Rowe nuclear plant in Massachusetts. NRC`s proposed changes would establish a screening criterion for both the rate of irradiation embrittlement or RPV materials and fracture toughness, above which a plant cannot continue to be operated safely.

Newman, P.

1994-10-13T23:59:59.000Z

215

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

Science Conference Proceedings (OSTI)

During certain phases of a severe accident in a pressurized water reactor (PWR), the core becomes uncovered and steam carries heat to the steam generators through natural circulation. For PWR's with U-tube steam generators and loop seals filled with water, a counter current flow pattern is established in the hot leg. This flow pattern has been experimentally observed and has been predicted using computational fluid dynamics (CFD). Predictions of severe accident behavior are routinely carried out using severe accident system analysis codes such as SCDAP/RELAP5 or MELCOR. These codes, however, were not developed for predicting the three-dimensional natural circulation flow patterns during this phase of a severe accident. CFD, along with a set of experiments at 1/7. scale, have been historically used to establish the flow rates and mixing for the system analysis tools. One important aspect of these predictions is the counter current flow rate in the nearly 30 inch diameter hot leg between the reactor vessel and steam generator. This flow rate is strongly related to the amount of energy that can be transported away from the reactor core. This energy transfer plays a significant role in the prediction of core failures as well as potential failures in other reactor coolant system piping. CFD is used to determine the counter current flow rate during a severe accident. Specific sensitivities are completed for parameters such as surge line flow rates, hydrogen content, as well as vessel and steam generator temperatures. The predictions are carried out for the reactor vessel upper plenum, hot leg, a portion of the surge line, and a steam generator blocked off at the outlet plenum. All predictions utilize the FLUENT V6 CFD code. The volumetric flow in the hot leg is assumed to be proportional to the square root of the product of normalized density difference, gravity, and hydraulic diameter to the 5. power. CFD is used to determine the proportionality constant in the range from 0.11 to 0.13 and termed a discharge coefficient. The value is relatively unchanged for typical surge line flow rates as well as the hydrogen content in the flow. Over a significant range of expected temperature differences for the steam generator and reactor vessel upper plenum, the discharge coefficient also remained consistent. The discharge coefficient is a suitable model for determining the hot leg counter current flow rates during this type of severe accident. (author)

Boyd, Christopher F. [United States Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

2006-07-01T23:59:59.000Z

216

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

DOE Green Energy (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

217

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

SciTech Connect

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

218

Reactor Core Assembly - HFIR Technical Parameters | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (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,

219

The removal and segmentation of the Yankee Rowe reactor vessel internals  

Science Conference Proceedings (OSTI)

A major element of the reactor decommissioning of the Rowe Yankee reactor was the segmentation and packaging of the reactor internals. PCI Energy Services, specializing in remote cutting, machining, and welding, performed this work under contract to Yankee Atomic Electric Company. Removal techniques are described.

Child, C.; McGough, M.; Smith, G. [Power Cutting Inc., Lake Bluff, IL (United States)

1995-12-31T23:59:59.000Z

220

Bobbin-Tool Friction-Stir Welding of Thick-Walled Aluminum Alloy Pressure Vessels  

SciTech Connect

It was desired to assemble thick-walled Al alloy 2219 pressure vessels by bobbin-tool friction-stir welding. To develop the welding-process, mechanical-property, and fitness-for-service information to support this effort, extensive friction-stir welding-parameter studies were conducted on 2.5 cm. and 3.8 cm. thick 2219 Al alloy plate. Starting conditions of the plate were the fully-heat-treated (-T62) and in the annealed (-O) conditions. The former condition was chosen with the intent of using the welds in either the 'as welded' condition or after a simple low-temperature aging treatment. Since preliminary stress-analyses showed that stresses in and near the welds would probably exceed the yield-strength of both 'as welded' and welded and aged weld-joints, a post-weld solution-treatment, quenching, and aging treatment was also examined. Once a suitable set of welding and post-weld heat-treatment parameters was established, the project divided into two parts. The first part concentrated on developing the necessary process information to be able to make defect-free friction-stir welds in 3.8 cm. thick Al alloy 2219 in the form of circumferential welds that would join two hemispherical forgings with a 102 cm. inside diameter. This necessitated going to a bobbin-tool welding-technique to simplify the tooling needed to react the large forces generated in friction-stir welding. The bobbin-tool technique was demonstrated on both flat-plates and plates that were bent to the curvature of the actual vessel. An additional issue was termination of the weld, i.e. closing out the hole left at the end of the weld by withdrawal of the friction-stir welding tool. This was accomplished by friction-plug welding a slightly-oversized Al alloy 2219 plug into the termination-hole, followed by machining the plug flush with both the inside and outside surfaces of the vessel. The second part of the project involved demonstrating that the welds were fit for the intended service. This involved determining the room-temperature tensile and elastic-plastic fracture-toughness properties of the bobbin-tool friction-stir welds after a post-weld solution-treatment, quenching, and aging heat-treatment. These mechanical properties were used to conduct fracture-mechanics analyses to determine critical flaw sizes. Phased-array and conventional ultrasonic non-destructive examination was used to demonstrate that no flaws that match or exceed the calculated critical flaw-sizes exist in or near the friction-stir welds.

Dalder, E C; Pastrnak, J W; Engel, J; Forrest, R S; Kokko, E; Ternan, K M; Waldron, D

2007-06-06T23:59:59.000Z

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

Materials Reliability Program: Pressurized Water Reactor Internals Aging Management Program Development Template (MRP-342)  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) has completed and published guidance for managing the effects of aging degradation in pressurized water reactor (PWR) internals. The initial version of this report, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0), was submitted to the staff of the U. S. Nuclear Regulatory Commission (NRC) ...

2012-10-23T23:59:59.000Z

222

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

223

An experimental study of assessment of weld quality on fatigue reliability analysis of a nuclear pressure vessel  

SciTech Connect

The steam generator in a PWR primary coolant system is one of the pieces of equipment made in China for the Qinshan nuclear power plant, Zhejiang. It is a crucial unit belonging to the category of nuclear pressure vessel. The purpose of this research work is to carry out an examination of the weld quality of the steam generator under fatigue loading and to assess its reliability by using experimental results of a fatigue test of the nuclear pressure vessel steel S-271 (Chinese Standard) and of qualified tests of welded seams of a simulated prototype of the bottom closure head of the steam generator. A guarantee of weld quality is proposed as the quality assurance of safety for the China National Nuclear Safety Supervision Bureau. The results of the reliability analysis reported in this work can be taken as supplementary material for a Probabilistic Risk Assessment (PRA) of the Qinshan nuclear power plant. According to the requirement of Provision 2-1500 CYCLIC TESTING, ASME Boiler and Pressure Vessel Code, Section 3, Rules for Construction of Nuclear Power Plant Components, a simulated prototype of the bottom closure head of the steam generator was made in this work for the qualified tests. Qualified tests with small sample size present a problem which is difficult to solve in reliability analysis, and are therefore of interest. Here, the authors offer proposals attempting to solve this problem.

Dai, Shuho (Nanjing Inst. of Chemical Technology, Jiangsu (China). Dept. of Mechanical Engineering)

1993-11-01T23:59:59.000Z

224

BWRVIP-60-A: BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in th e BWR Environment  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals materials issues. This report provides a methodology for assessing crack growth in BWR low alloy steel pressure vessels and nozzles. A previous version of this report was published as BWRVIP-60 (TR-108709). This report (BWRVIP-60-A) incorporates the U.S. Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) and ot...

2003-06-09T23:59:59.000Z

225

CO/sub 2/ welding used to attach inspection manway to NASA hydrogen pressure vessel  

SciTech Connect

Welding of inspection manway for internal survey of a gaseous hydrogen storage vessel is described. Pre-welding activities are reviewed, along with welding operations, and in-process welding control. (JRD)

Palmer, G.; Conklin, D.

1976-09-01T23:59:59.000Z

226

Accident management for indian pressurized heavy water reactors  

Science Conference Proceedings (OSTI)

Indian nuclear power program as of now is mainly based on Pressurized Heavy Water Reactors (PHWRs). Operating Procedures for normal power operation and Emergency Operating Procedures for operational transients and accidents within design basis exist for all Indian PHWRs. In addition, on-site and off-site emergency response procedures are also available for these NPPs. The guidelines needed for severe accidents mitigation are now formally being documented for Indian PHWRs. Also, in line with International trend of having symptom based emergency handling, the work is in advanced stage for preparation of symptom-based emergency operating procedures. Following a plant upset condition; a number of alarms distributed in different information systems appear in the control room to aid operator to identify the nature of the event. After identifying the event, appropriate intervention in the form of event based emergency operating procedure is put into use by the operating staff. However, if the initiating event cannot be unambiguously identified or after the initial event some other failures take place, then the selected event based emergency operating procedure will not be optimal. In such a case, reactor safety is ensured by monitoring safety functions (depicted by selected plant parameters grouped together) throughout the event handling so that the barriers to radioactivity release namely, fuel and fuel cladding, primary heat transport system integrity and containment remain intact. Simultaneous monitoring of all these safety functions is proposed through status trees and this concept will be implemented through a computer-based system. For beyond design basis accidents, event sequences are identified which may lead to severe core damage. As part of this project, severe accident mitigation guidelines are being finalized for the selected event sequences. The paper brings out the details of work being carried out for Indian PHWRs for symptom based event handling and severe accident management. (authors)

Hajela, S.; Grover, R.; Ghadge, S.G.; Bajaj, S.S. [Directorate of Safety, Nuclear Power Corporation of India Limited Nabhikiya Urja Bhawan, Anushakti Nagar, Mumbai-400 094 (India)

2006-07-01T23:59:59.000Z

227

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

228

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

229

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor  

SciTech Connect

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

Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

230

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

SciTech Connect

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

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

2012-04-01T23:59:59.000Z

231

Development of Heat Transfer Enhancement Techniques for ExternalCooling of an Advanced Reactor Vessel.  

E-Print Network (OSTI)

??Nucleate boiling is a well-recognized means for passively removinghigh heat loads (up to 106 W/m2) generated by amolten reactor core under severe accident conditions whilemaintaining… (more)

Yang, Jun

2005-01-01T23:59:59.000Z

232

TECHNICAL BASIS AND APPLICATION OF NEW RULES ON FRACTURE CONTROL OF HIGH PRESSURE HYDROGEN VESSEL IN ASME SECTION VIII, DIVISION 3 CODE  

DOE Green Energy (OSTI)

As a part of an ongoing activity to develop ASME Code rules for the hydrogen infrastructure, the ASME Boiler and Pressure Vessel Code Committee approved new fracture control rules for Section VIII, Division 3 vessels in 2006. These rules have been incorporated into new Article KD-10 in Division 3. The new rules require determining fatigue crack growth rate and fracture resistance properties of materials in high pressure hydrogen gas. Test methods have been specified to measure these fracture properties, which are required to be used in establishing the vessel fatigue life. An example has been given to demonstrate the application of these new rules.

Rawls, G

2007-04-30T23:59:59.000Z

233

EPRI Materials Management Matrix Project: Advanced Light-Water Reactor - Pressurized Water Reactor Degradation Matrix - Revision 1  

Science Conference Proceedings (OSTI)

The Advanced Light Water Reactor - Pressurized Water Reactor Degradation Matrix (ALWR PWR DM) is an integral piece of the Electric Power Research Institutes (EPRIs) Materials Management Matrix (MMM) initiative for ALWR designs. The MMM provides a tool to assist the industry in proactive identification and consideration of materials issues and mitigation/management opportunities from the design phase through component fabrication and plant construction to operations and maintenance.

2010-09-22T23:59:59.000Z

234

Materials Reliability Program: Background of Reactor Vessel Material Surveillance Programs and Recommended Data Validation Protocol for the Radiation Embrittlement Archive Project (REAP) Database (MRP-369)  

Science Conference Proceedings (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) staff is currently developing an updated, comprehensive reactor vessel surveillance database as a replacement for the Power Reactor Embrittlement Database Version 3 (PR-EDB). The new database—called “Radiation Embrittlement Archive Project,” or REAP—will be available to the public via the Internet. The REAP database contains essential information on the materials and test results from surveillance capsules in U.S. plants; these ...

2013-07-25T23:59:59.000Z

235

The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors  

E-Print Network (OSTI)

Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the ...

Kamal, Altamash

1982-01-01T23:59:59.000Z

236

Design strategies for optimizing high burnup fuel in pressurized water reactors  

E-Print Network (OSTI)

This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

Xu, Zhiwen, 1975-

2003-01-01T23:59:59.000Z

237

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

238

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

239

Analysis of strategies for improving uranium utilization in pressurized water reactors  

E-Print Network (OSTI)

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

Sefcik, Joseph A.

1981-01-01T23:59:59.000Z

240

Optimal design of a high pressure organometallic chemical vapor deposition reactor  

Science Conference Proceedings (OSTI)

A team composed of material scientists, physicists, and applied mathematicians have used computer simulations as a fundamental design tool in developing a new prototype High Pressure Organometallic Chemical Vapor Deposition (HPOMCVD) reactor for use ...

K. J. Bachmann; H. T. Banks; C. Höpfner; G. M. Kepler; S. Lesure; S. D. Mccall; J. S. Scroggs

1999-04-01T23:59:59.000Z

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

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

242

2012 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guideline Revision 7  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) periodically updates its Pressurized Water Reactor (PWR) water chemistry guidelines as new information becomes available and as required by Nuclear Energy Institute (NEI) 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). An industry review committee meeting in September 2012 determined that a revision of the 2009 version of EPRI’s Pressurized Water Reactor Secondary ...

2012-12-21T23:59:59.000Z

243

2013 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guideline Revision 7  

Science Conference Proceedings (OSTI)

As required by Nuclear Energy Institute (NEI) 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues), the Electric Power Research Institute (EPRI) periodically updates its Pressurized Water Reactor (PWR) water chemistry guidelines when new information becomes available. An industry review committee meeting in June 2013 determined that a revision of the 2009 version of EPRI’s Pressurized Water Reactor Secondary Water ...

2013-10-18T23:59:59.000Z

244

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

245

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  

Science Conference Proceedings (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

246

Optimization of hydride fueled pressurized water reactor cores  

E-Print Network (OSTI)

This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

Shuffler, Carter Alexander

2004-01-01T23:59:59.000Z

247

Innovative fuel designs for high power density pressurized water reactor  

E-Print Network (OSTI)

One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

Feng, Dandong, Ph. D. Massachusetts Institute of Technology

2006-01-01T23:59:59.000Z

248

Life of Plant Activity Estimates for a Nominal 1000 MWe Pressurized Water Reactor and Boiling Water Reactor  

Science Conference Proceedings (OSTI)

Decommissioning nuclear power plant and disposal site managers must understand the radioactive source term of a nuclear power plant to effectively manage disposition of these materials. This study estimates the radioactive source term from nominal 1000 MWe pressurized water and boiling water reactors to support decisions related to radioactive waste storage, processing, and disposal through decommissioning.BackgroundThis study examines the radionuclide ...

2012-12-05T23:59:59.000Z

249

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

250

Experimental and Numerical Investigation of Boron Dilution Transients in Pressurized Water Reactors  

SciTech Connect

Within the pressurized water reactor (PWR) safety analyses, attention has increasingly focused in recent years on boron dilution events that could potentially lead to reactivity transients. Mixing of the low-boron water with the ambient coolant of higher boron content provides an important mitigation mechanism before the low-boron water enters the core.Experimental support is needed to validate the computational tools to be applied to analyze the mixing of the low-boron water. Experiments were performed in the three test facilities - the Upper Plenum Test Facility (UPTF), the Primaerkreislauf (PKL), and the Rossendorf coolant mixing model (ROCOM) - in Germany.The relevant PKL and UPTF tests were focused on small-break loss-of-coolant accident (SBLOCA) scenarios with reflux-condenser mode and restart of natural circulation. The two test facilities represent a typical western-type PWR and are/were operated by Siemens/KWU now Framatome ANP in Germany. While the restart of natural circulation was investigated in the PKL system test facility (volume 1:145, height 1:1), the UPTF experiments dealt with the mixing of water flows with different boron concentration in the cold legs, reactor pressure vessel (RPV) downcomer, and the lower plenum (all these components were full-scale models).The results from the PKL test facility demonstrate that in case of a postulated SBLOCA with reflux condensation phase, natural circulation does not start up simultaneously in all loops. This means that slugs of condensate, which might have accumulated in the pump seal during reflux-condenser mode of operation, would reach the RPV at different points in time. The UPTF tests showed an almost ideal mixing of water flows with different boron concentration in the RPV downcomer.The ROCOM test facility has been built in a linear scale of 1:5 for the investigation of coolant mixing phenomena in a wide range of flow conditions in the RPV of the German KONVOI-type PWR. The test results presented are focused on the mixing of a slug of deborated water during the startup of the first reactor coolant pump. Based on experimentally determined pulse responses, a semianalytical model for the description of coolant mixing inside the KONVOI RPV has been developed. Calculations for a presumed boron dilution event during the startup of the first reactor coolant pump have been carried out by means of the semianalytical model and independently by means of the computational fluid dynamics code CFX-4. The semianalytical model is able to describe the time dependent behavior of the deboration front at each fuel element position in a good agreement with the experiment. All main mixing effects, observed in the experiment, are also reproduced by the CFX calculation.

Hertlein, Roland J. [Framatome ANP GmbH (France); Umminger, Klaus [Framatome ANP GmbH (France); Kliem, Soeren [Forschungszentrum Rossendorf e.V. (Germany); Prasser, Horst-Michael [Forschungszentrum Rossendorf e.V. (Germany); Hoehne, Thomas [Forschungszentrum Rossendorf e.V. (Germany); Weiss, Frank-Peter [Forschungszentrum Rossendorf e.V. (Germany)

2003-01-15T23:59:59.000Z

251

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

252

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

SciTech Connect

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

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

2003-04-01T23:59:59.000Z

253

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

DOE Green Energy (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

254

Pressure suppression containment system  

DOE Patents (OSTI)

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto.

Gluntz, Douglas M. (San Jose, CA); Townsend, Harold E. (San Jose, CA)

1994-03-15T23:59:59.000Z

255

Pressure suppression containment system  

DOE Patents (OSTI)

A pressure suppression containment system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The wetwell pool includes a plenum for receiving the non-condensable gas carried with steam from the drywell following a loss-of-coolant-accident (LOCA). The wetwell plenum is vented to a plenum above the GDCS pool following the LOCA for suppressing pressure rise within the containment vessel. A method of operation includes channeling steam released into the drywell following the LOCA into the wetwell pool for cooling along with the non-condensable gas carried therewith. The GDCS pool is then drained by gravity, and the wetwell plenum is vented into the GDCS plenum for channeling the non-condensable gas thereto. 6 figures.

Gluntz, D.M.; Townsend, H.E.

1994-03-15T23:59:59.000Z

256

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

SciTech Connect

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

257

Modeling of Late Blooming Phases and Precipitation Kinetics in Aging Reactor Pressure Vessel (RPV) Steels  

Science Conference Proceedings (OSTI)

The principle work at the atomic scale is to develop a predictive quantitative model for the microstructure evolution of RPV steels under thermal aging and neutron radiation. We have developed an AKMC method for the precipitation kinetics in bcc-Fe, with Cu, Ni, Mn and Si being the alloying elements. In addition, we used MD simulations to provide input parameters (if not available in literature). MMC simulations were also carried out to explore the possible segregation/precipitation morphologies at the lattice defects. First we briefly describe each of the simulation algorithms, then will present our results.

Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

2013-09-01T23:59:59.000Z

258

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

SciTech Connect

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

259

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

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

Treshow, M.

1959-02-10T23:59:59.000Z

260

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1958-08-19T23:59:59.000Z

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to obtain the most current and comprehensive results.


261

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

NLE Websites -- All DOE Office Websites (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 *

262

Equations for gas releasing process from pressurized vessels in ODH evaluation  

Science Conference Proceedings (OSTI)

The evaluation of Oxygen Deficiency Hazard (ODH) is a critical part in the design of any cryogenic system. The high-pressure gas tank or low-temperature liquid container that contain asphyxiated fluid could be the sources to bring about the oxygen deficiency hazard. In the evaluation of ODH

L. X. Jia; L. Wang

2002-01-01T23:59:59.000Z

263

An inverted hydride-fueled pressurized water reactor concept  

E-Print Network (OSTI)

Previous studies conducted at MIT showed that power performance of typical pin geometry PWRs are limited by three main constraints: core pressure drop, critical heat flux (CHF) and fretting phenomena of the fuel rods against ...

Ferroni, Paolo, Ph. D. Massachusetts Institute of Technology

2010-01-01T23:59:59.000Z

264

Radiant vessel auxiliary cooling system  

DOE Patents (OSTI)

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Germer, John H. (San Jose, CA)

1987-01-01T23:59:59.000Z

265

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

Science Conference Proceedings (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

266

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

DOE Green Energy (OSTI)

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

Ginsberg, T.; Tutu, N.K.

1988-01-01T23:59:59.000Z

267

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

Science Conference Proceedings (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

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

2007-06-30T23:59:59.000Z

268

Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references  

SciTech Connect

Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

Grotke, G.E.

1980-04-01T23:59:59.000Z

269

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

DOE Patents (OSTI)

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

Hill, P.R.

1994-12-27T23:59:59.000Z

270

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

DOE Patents (OSTI)

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

Hill, Paul R. (Tucson, AZ)

1994-01-01T23:59:59.000Z

271

Light Water Reactor Sustainability Technical Documents | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

272

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables—Revision 2 (MRP-205)  

Science Conference Proceedings (OSTI)

Ongoing issues related to the degradation of pressurized water reactor (PWR) nuclear steam supply system (NSSS) components have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and associated Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) and Steam Generator Management Program (SGMP) requirements.

2010-10-27T23:59:59.000Z

273

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

Science Conference Proceedings (OSTI)

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

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

2009-04-01T23:59:59.000Z

274

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

Science Conference Proceedings (OSTI)

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

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

2000-02-01T23:59:59.000Z

275

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

SciTech Connect

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

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

2000-02-01T23:59:59.000Z

276

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

Science Conference Proceedings (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

277

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

NLE Websites -- All DOE Office Websites (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

278

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

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

Treshow, M.

1958-11-18T23:59:59.000Z

279

REACTOR  

DOE Patents (OSTI)

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

Roman, W.G.

1961-06-27T23:59:59.000Z

280

Characteristics of Spent Fuel from Plutonium Disposition Reactors. Vol. 3: A Westinghouse Pressurized-Water Reactor Design  

Science Conference Proceedings (OSTI)

This report discusses the results of a simulation study involving the burnup of mixed-oxide (MOX) fuel in a Westinghouse pressurized-water reactor (PWR). The MOX was composed of uranium and plutonium oxides, where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program and considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition, the activities, and the decay heat, together with the gamma and neutron dose rates are discussed for the spent fuel. For the steady-state situation involving this PWR burning MOX fuel, two burn histories are reported. In one case, an assembly is burned in the reactor for two cycles, and in the second case and assembly is burned for three cycles. Furthermore, assemblies containing wet annular burnable absorbers (WABAs) and assemblies that do not contain WABAs are considered in all cases. The two-cycle cases have a burnup of 35 GWd/t, and the three-cycle cases have a burnup of 52.5 GWd/t.

Murphy, B.D.

1997-07-01T23:59:59.000Z

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

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

282

Materials Reliability Program: Safety Evaluation for Boric Acid Wastage of PWR Reactor Vessel Bottom Heads Due to Bottom-Mounted Noz zle Leakage (MRP-167)  

Science Conference Proceedings (OSTI)

This safety assessment addresses one of the potential safety issues associated with aging degradation of reactor vessel bottom head penetrations: bottom mounted nozzles (BMNs). Specifically, this report evaluates the concern that BMN leakage due to primary water stress corrosion cracking (PWSCC) of the Alloy 600 nozzle and/or Alloy 82/182 J-groove attachment weld could lead to significant wastage of the low-alloy steel head shell material due to concentration of the boric acid present in the leaking prim...

2008-07-02T23:59:59.000Z

283

Detailed Analysis of In-Vessel Melt Progression in the Loss of Coolant Accident of OPR1000  

Science Conference Proceedings (OSTI)

An in-vessel severe accident progression has been analyzed to generate the basic data for an evaluation of the in-vessel severe accident management strategies and to identify the thermal hydraulic condition of the reactor vessel and the damage state of the in-vessel materials at a reactor vessel failure by using the SCDAP/RELAP5/MOD3.3 computer code during the Loss Of Coolant Accident (LOCA) without the Safety Injection (SI) of the OPR (Optimized Pressurize Reactor) 1000. Best estimate calculation of the small break LOCAs of 1.35 inch and 2 inch, the medium break LOCAs of 3 inch and a 4.28 inch, and a large break LOCA of 9.8 inch without the SI have been performed from a transient initiation to a reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that in all the transients, approximately 30-40 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of a reactor vessel failure. In the small and large break LOCAs, the reactor vessel failed at an early time of approximately 70-110 minutes after the transients were initiated. Since the Safety Injection Tanks (SITs) were actuated effectively in the medium break LOCAs, the reactor vessel failed at a later time of approximately 200-400 minutes after the transients were initiated. At the time of a reactor vessel failure, approximately 45-55 % of the fuel rod cladding was oxidized in the small and medium break LOCAs. However, approximately 20 % of the fuel rod cladding was oxidized because of a coolant loss through the break in the large break LOCA of the OPR1000. (authors)

Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2006-07-01T23:59:59.000Z

284

2011 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guidelines -- Revision 7  

Science Conference Proceedings (OSTI)

EPRI periodically updates its PWR water chemistry guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). An industry review committee meeting in September 2011 determined that a revision of the 2009 version of EPRI's Pressurized Water Reactor Secondary Water Chemistry Guidelines is not warranted at this time, nor is interim guidance required.

2011-12-15T23:59:59.000Z

285

Evaluation of Fuel Clad Corrosion Product Deposits and Circulating Corrosion Products in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Many pressurized water reactors (PWRs) have experienced negative consequences resulting from build-up of corrosion product deposits (crud) on fuel cladding. The negative consequences include unplanned shifts in core power (axial offset anomaly, or AOA), fuel cladding failure, anomalous shutdown chemistry, and elevated ex-core radiation fields. These problems have grown more common as PWRs have moved toward higher 235U enrichments and higher duty cores needed for extended cycle operation. This report expl...

2004-12-08T23:59:59.000Z

286

Below Core Plate Inspections for Foreign Material in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

During the period 2000–2011, 24 U.S. pressurized water reactors (PWRs) experienced fuel cladding failures by debris-induced fretting over 38 cycles (source: EPRI Fuel Reliability Database). In 2006, U.S. chief nuclear officers endorsed the “Zero-by- Ten” initiative with the stated goal of reducing all fuel failures to zero by 2010. This effort was very successful, but some failure mechanisms continue to occur. Currently, the mechanism most prominent in PWRs is grid to rod fretting. ...

2012-11-07T23:59:59.000Z

287

Materials Reliability Program: Pressurized Water Reactor Issue Management Tables - Revision 3 (MRP-205)  

Science Conference Proceedings (OSTI)

Nuclear utilities continue to face a number of ongoing issues related to degradation of pressurized water reactor (PWR) nuclear steam supply system (NSSS) components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) issues and associated EPRI Materials Reliability Program (MRP) and Steam Generator Management Program (SGMP) requirements.BackgroundA comprehensive, integrated ...

2013-12-03T23:59:59.000Z

288

Minimization of Pressurized Water Reactor Radiation Fields through Fuel Deposit Engineering: Deposit Property Evaluation and Optimization  

Science Conference Proceedings (OSTI)

The purpose of this report is to provide an initial assessment of the options for modification of pressurized water reactor (PWR) primary side corrosion product deposits (crud) to minimize the incorporation of activated crud into out-of-core surfaces, thus reducing the intensity of out-of-core radiation fields. This report summarizes the current knowledge of PWR fuel crud characteristics, including crystallographic structure (crystal habits), and buildup mechanisms. The report also reviews the ...

2013-11-11T23:59:59.000Z

289

Pressurized Water Reactor Zinc Application: Data Analysis and Evaluation of Primary Chemistry Responses  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) Pressurized Water Reactor Zinc Application Users Group (PWR ZUG) facilitates and improves the use of zinc injection in PWR primary coolant systems by assisting in the evaluation of zinc injection performance; documentation of lessons learned; communication of information on zinc injection qualification, monitoring, and operating experience; and review of zinc application effectiveness regarding primary water stress corrosion cracking (PWSCC) and radiation fiel...

2010-08-15T23:59:59.000Z

290

Materials Reliability Program: Risk Assessment of ASME Section XI Appendix G Pressure-Temperature (P-T) Limit Curve Methodologies (M RP-368)  

Science Conference Proceedings (OSTI)

This report presents the results of an assessment of the conditional probability of reactor pressure vessel (RPV) failure in pressurized water reactors (PWRs) when normal RPV heatup and cooldown occur along operational constraint boundaries. These boundaries are defined by the maximum allowable pressures determined from regulatory requirements, the evaluation procedures in Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Appendix G), ...

2013-12-09T23:59:59.000Z

291

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)  

Science Conference Proceedings (OSTI)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Roh, Gyuhong; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

2004-06-15T23:59:59.000Z

292

BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan  

Science Conference Proceedings (OSTI)

This report describes the boiling water reactor (BWR) Integrated Surveillance Program (ISP). Based on recommendations from BWR Vessel and Internals Project (BWRVIP) utilities, it was concluded that combining all separate BWR surveillance programs into a single integrated program would be beneficial. In the integrated program, representative materials chosen for a specific reactor pressure vessel (RPV) can be materials from another plant surveillance program or other source that better represents the ...

2012-10-01T23:59:59.000Z

293

US EPR Tests Performed to confirm the Mechanical and Hydraulic Design of the Vessel Internals  

SciTech Connect

The EPR is an Evolutionary high-Power Reactor which is based on the best French and German experience of the past twenty years in plant design construction and operation. In the present detailed engineering phase of the plant under construction in Finland (Okiluoto 3) or scheduled in France (Flamanville 3), a few actions are still ongoing mainly to complement equipment validation files. Design and validation of the main EPR components were performed within Framatome ANP's engineering teams and its two Technical Centers located in France and Germany, which develop state of the art methods in the field of thermo hydraulic testing. The Reactor Pressure Vessel internals are mainly derived from components already implemented on presently operating plants, but they differ in some features from the design used in French N4 or German Konvoi. The aim of this paper is to present the tests performed to confirm the hydraulic and mechanical design of the EPR vessel internals. - Four different mock-ups are presented to illustrate these tests: - JULIETTE for the reactor pressure vessel lower internals; - ROMEO for the reactor pressure vessel upper internals; - MAGALY for the design of the skeleton-type control rod guide assembly; - HYDRAVIB for the vibratory response of the reactor pressure vessel lowers internals. (authors)

Dolleans, Philippe; Chambrin, Jean-Luc; Muller, Thierry [FRAMATOME ANP, Tour AREVA 1 place de la Coupole, 92084 PARIS La D ense (France)

2006-07-01T23:59:59.000Z

294

Pressure suppression system  

DOE Patents (OSTI)

A pressure suppression system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and an enclosed gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel. The GDCS pool includes a plenum for receiving through an inlet the non-condensable gas carried with steam from the drywell following a loss-of-coolant accident (LOCA). A condenser is disposed in the GDCS plenum for condensing the steam channeled therein and to trap the non-condensable gas therein. A method of operation includes draining the GDCS pool following the LOCA and channeling steam released into the drywell following the LOCA into the GDCS plenum for cooling along with the non-condensable gas carried therewith for trapping the gas therein.

Gluntz, Douglas M. (San Jose, CA)

1994-01-01T23:59:59.000Z

295

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

SciTech Connect

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

Rachamin, R.; Fridman, E. [Reactor Safety Div., Inst. of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf, POB 51 01 19, 01314 Dresden (Germany); Galperin, A. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, POB 653, Beer Sheva 84105 (Israel)

2012-07-01T23:59:59.000Z

296

Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation  

E-Print Network (OSTI)

The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

Buongiorno, J.

297

Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation  

E-Print Network (OSTI)

The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

Buongiorno, Jacopo, 1971-

2001-01-01T23:59:59.000Z

298

In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues  

Science Conference Proceedings (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 Advanced 600 MWe Pressurized Water Reactor (PWR) designed by Westinghouse (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 Light Water Reactors (LWRs). However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors (up to 1500 MWe) without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high power thermal reactors.

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

2008-03-01T23:59:59.000Z

299

Risk contribution from low power and shutdown of a pressurized water reactor  

Science Conference Proceedings (OSTI)

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs.

Chu, T.L.; Pratt, W.T.

1997-04-01T23:59:59.000Z

300

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.

Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

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.


301

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

302

Main results of study on the interaction between the corium melt and steel in the VVER-1000 reactor vessel during a severe accident performed under the MASCA project  

Science Conference Proceedings (OSTI)

The interactions that take place in the corium melt in the reactor vessel in the case of a severe accident at a nuclear power plant were investigated in accordance with the MASCA international program. Results of the interaction between the oxide melt and iron (steel), partition of the main components [U, Zr, Fe (stainless steel)] between the oxide and the metal phases of the melt, partition of low-volatile simulators of fission products between the phases of the stratified core melt pool, and impact of the oxidizing atmosphere on the melt stratification are presented. The results obtained were used for prediction of thermodynamic properties of the melts belonging to the U-Zr-Fe-O system.

Asmolov, V. G.; Zagryazkin, V. N.; Tsurikov, D. F. [Russian Research Center Kurchatov Institute (Russian Federation); Vishnevsky, V. Yu.; D'yakov, Ye. K.; Kotov, A. Yu.; Repnikov, V. M. [Research Institute Scientific Production Association Lutch (Russian Federation)

2010-12-15T23:59:59.000Z

303

Emergency heat removal system for a nuclear reactor  

DOE Patents (OSTI)

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Dunckel, Thomas L. (Potomac, MD)

1976-01-01T23:59:59.000Z

304

Materials Reliability Program: Input for Pressurized Thermal Shock Rulemaking (MRP-248)  

Science Conference Proceedings (OSTI)

The Pressurized Thermal Shock (PTS) rule addresses the risk of a nuclear power plant reactor vessel failing due to propagation of a crack through the vessel wall. If a plant has an emergency cool-down event that superimposes a large thermal transient stress on a large pressure stress in the presence of a pre-existing flaw, it is possible that a crack could initiate and propagate through the vessel wall. The resistance of a vessel to crack initiation and propagation declines as the vessel ages. As a resul...

2008-12-23T23:59:59.000Z

305

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980  

SciTech Connect

Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

Not Available

1981-08-01T23:59:59.000Z

306

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

SciTech Connect

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

307

Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station  

SciTech Connect

A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies.

Carbajo, J.J.

1995-06-01T23:59:59.000Z

308

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

DOE Green Energy (OSTI)

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

Curran, H.J.; Pitz, W.J.; Westbrook, C.K. [Lawrence Livermore National Lab., CA (United States); Callahan, C.V.; Dryer, F.L. [Princeton Univ., Areospace Engineering. NJ (United States)

1998-01-01T23:59:59.000Z

309

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

310

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

SciTech Connect

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

Gagner, Renata; Lafitte, Helene; Dormeau, Pascal [Framatome ANP, Tour Areva - 1, place de la Coupole - 92084 Paris La Defense (France); Stoudt, Roger H. [Framatome ANP, Lynchburg - 3315 Old Forest Road, Lynchburg, VA 24501 (United States)

2004-07-01T23:59:59.000Z

311

CIVILIAN POWER REACTOR PROGRAM. PART I. SUMMARY OF TECHNICAL AND ECONOMIC STATUS AS OF 1960. HEAVY WATER-MODERATED POWER REACTORS  

SciTech Connect

A revision of section 8 of TID-8516, Part I, is presented. The reactor concept which is presented is a change from the former design, a pressurized- pressure vessel-indirect cycle plant to a boiling-pressure tubedirect cycle plant. A description of the plant and a summary of characteristics for the 110- and 325- Mwe systems are given. (J.R.D.)

Hutton, J.H.; Davis, S.A.; Graves, C.C.; Duffy, J.G. comps.

1960-08-19T23:59:59.000Z

312

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

Science Conference Proceedings (OSTI)

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

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

1981-09-15T23:59:59.000Z

313

Nonlinear response of vessel walls due to short-time thermomechanical loading  

Science Conference Proceedings (OSTI)

Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented.

Pfeiffer, P.A.; Kulak, R.F.

1994-06-01T23:59:59.000Z

314

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

Science Conference Proceedings (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

315

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  

NLE Websites -- All DOE Office Websites (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

316

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

directly in the steam generator. Smaller corrections arefeed-water to the steam generator (PWR) or reactor vessel (secondary side of the steam generator or reactor vessel. The

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

317

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

318

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

319

Start-up control system and vessel for LMFBR  

DOE Patents (OSTI)

A reflux condensing start-up system comprises a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

Durrant, Oliver W. (Akron, OH); Kakarala, Chandrasekhara R. (Clinton, OH); Mandel, Sheldon W. (Galesburg, IL)

1987-01-01T23:59:59.000Z

320

Start-up control system and vessel for LMFBR  

DOE Patents (OSTI)

A reflux condensing start-up system includes a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

Durrant, Oliver W. (Akron, OH); Kakarala, Chandrasekhara R. (Clinton, OH); Mandel, Sheldon W. (Galesburg, IL)

1987-01-01T23:59:59.000Z

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.


321

Nuclear reactor  

DOE Patents (OSTI)

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

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

322

BWRVIP-167NP, Rev. 3: Boiling Water Reactor Issue Management Tables  

Science Conference Proceedings (OSTI)

Nuclear utilities continue to face a number of ongoing issues related to degradation of boiling water reactor (BWR) pressure vessels, reactor internals, and American Society of Mechanical Engineers (ASME) Class 1 piping components. These issues have resulted in the need for a summary tool to assist in prioritizing and addressing research and development (R&D) gaps and BWR Vessel and Internals Project (BWRVIP) requirements. The BWR Issue Management Tables in the report are living documents that ...

2013-08-23T23:59:59.000Z

323

Enhancements of a Combustion Vessel to Determine Laminar Flame Speeds of Hydrocarbon Blends with Helium Dilution at Elevated Temperatures and Pressures  

E-Print Network (OSTI)

Fuel flexibility in gas turbines is of particular importance because of the main fuel source, natural gas. Blends of methane, ethane, and propane are big constituents in natural gas and consequently are of particular interest. With this level of importance comes the need for baseline data such as laminar flame speed of said fuels. While flame speeds at standard temperature and pressure have been extensively studied in the literature, experimental data at turbine-like conditions are still lacking currently. This thesis discusses the theory behind laminar flames; new data acquisition techniques; temperature and pressure capability improvements; measured flame speeds; and a discussion of the results including stability analysis. The measured flame speeds were those of methane, ethane, and propane fuel blends, as well as pure methane, at an elevated pressure of 5 atm and temperatures of 298 and 473 K, using a constant-volume, cylindrical combustion vessel. The current Aramco mechanism developed in conjunction with National University of Ireland Galway compared favorably with the data, while the literature data showed discrepancies at stoichiometric to rich conditions. An in-depth flame speed uncertainty analysis yielded a wide range of values from 0.5 cm/s to 21.5 cm/s. It is well known that high-pressure experiments develop flame instabilities when air is used as the oxidizer. In this study, the hydrodynamic instabilities were restrained by using a high diluent-to-oxygen ratio. The thermal-diffusive instabilities were inhibited by using helium as the diluent. To characterize this flame stability, the Markstein length and Lewis number were calculated for the presented conditions. The resultant positive Markstein lengths showed a low propensity of flame speed to flame stretch, while the larger-than-unity Lewis numbers showed the relatively higher diffusivity of helium to that of nitrogen.

Plichta, Drew

2013-05-01T23:59:59.000Z

324

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

SciTech Connect

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

Chodak, P. III

1996-05-01T23:59:59.000Z

325

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

NLE Websites -- All DOE Office Websites (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

326

Friction pressure drop measurements and flow distribution analysis for LEU conversion study of MIT Research Reactor  

E-Print Network (OSTI)

The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer. Recent studies on the conversion to low-enriched ...

Wong, Susanna Yuen-Ting

2008-01-01T23:59:59.000Z

327

Introduction of Thorium-Based Fuels in High Conversion Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Regular Technical Paper / Special Issue on the Symposium on Radiation Effects in Ceramic Oxide and Novel LWR Fuels / Fission Reactors

V. Vallet; B. Gastaldi; J. Politello; A. Santamarina; L. Van Den Durpel

328

Thermal energy storage using Prestressed Cast Iron Vessels (PCIV). Final report  

DOE Green Energy (OSTI)

The wide-spread application of thermal energy and high-pressure air storage to electric power generation has so far been hampered by the lack of large high-pressure storage vessels of reasonable cost. Welded steel vessels are too expensive for this purpose. However, the Prestressed Cast Iron Vessel (PCIV), developed as a nuclear reactor pressure vessel by Siempelkamp Giesserei KG of Krefeld, FRG, has the potential of complying with these requirements. Applications of the PCIV include: high-pressure air storage for the quick start-up of open cycle gas turbines; pressurized high-temperature sensible heat storage by means of solids with a gaseous heat transfer medium for closed cycle gas turbines of future solar power stations; and pressurized hot water storage for nuclear, solar, or coal-fired steam power plants, employing either separate peaking turbines or overloadable main turbine sets. A reference PCIV of 8000 m/sup 3/, 275/sup 0/C, with hot going walls and cold going tendons was developed, designed, and stress-analysed. A parametric study showed that pressures between 4 and 8 MPa and L/D ratios larger than 4 should be optimal. Cost of the reference vessel is about $10,000,000 or 33 to 50 $/kWh electric energy stored. Cost of peak power will be from 30 to 100 mills/kWh, depending on many parameters.

Gilli, P.V.; Beckmann, G.; Schilling, F.E.

1977-06-01T23:59:59.000Z

329

Application of the LEPRICON Methodology to the Arkansas Nuclear One, Unit 1 Reactor  

Science Conference Proceedings (OSTI)

Applying the new LEPRICON dosimetry methodology to irradiation data from the ANO-1 reactor greatly reduced uncertainties in pressure vessel fluence estimates. Accurate estimates will enable safety engineers to better predict reactor lifetimes and will eliminate the need for overly conservative safety margins.

1986-02-19T23:59:59.000Z

330

REACTOR FUEL WASTE DISPOSAL PROJECT PRESSURE-TEMPERATURE EFFECT ON SALT CAVITIES AND SURVEY OF LIQUEFIED PETROLEUM GAS STORAGE  

SciTech Connect

It is deemed feasible to store reactor fuel wastes in a salt dome cavity to a depth where the differential in pressure between the soil over-burden pressure and pressure of the fluid inside the cavity does not exceed 3000 psi, and the temperature is less than 400 deg F. Tests at pressure increments of 1000 psi were conducted on a 2" cylindrical cavity contained in a 6-in. long by 6-in. cylindrical salt core. Tests indicate that the cavity exhibited complete stability under pressures to 3000 psi and temperatures to 300 deg F. At temperatures of 100 to 400 deg F and pressures to 5000 psi continuous deformation of the cavity resulted. Initial movement of the salt was observed at all pressures. This was evidenced by vertical deformation and cavity size reduction. It was noted that a point of structural equilibrium was reached at lower temperatures when the pressure did not exceed 5000 psi. A literature study reveals that the most common type of cavity utilized in liquefied petroleum gas storage is either cylindrical or ellipsoidal. A few are pear or inverted cone shaped. There was no indication of leakage for cavities when pressure tested for as long as 72 hr. This indicates that the salt mass is not permeable under conditions of prevailing underground temperature and pressure. Salt specimens tested under atmospheric Pressure and temperature exhibited permeabilities of 0.1 to 0.2 millidarcys. The cost of completing underground storage cavities in salt masses is expected to be approximately 05 per barrel of storage space. (auth)

Brown, K.E.; Jessen, F.W.; Gloyna, E.F.

1959-01-15T23:59:59.000Z

331

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants  

Science Conference Proceedings (OSTI)

Technical Paper / Second Seminar on Accelerated Testing of Materials in Spent Nuclear Fuel and High-Level Waste Storage Systems / Fission Reactors

Aung Tharn Daing; Myung-Hyun Kim

332

F-8: Modeling of Mn-Ni-Si-Cu Precipitation in Reactor Pressure ...  

Science Conference Proceedings (OSTI)

Presentation Title, F-8: Modeling of Mn-Ni-Si-Cu Precipitation in Reactor .... Steels 316 and Comparison with the Rate Theory Model of a Multicomponent System.

333

Comparison of ALICE-II code predictions with SRI complex vessel experiments  

Science Conference Proceedings (OSTI)

Several complex vessel experiments on 1/20-scale models of the Clinch River Breeder Reactor Project (CRBR) were performed by SRI International to help evaluate the containment structural integrity subjected to HCDAs. Among these experiments SM-3 is a simple model which consists of a radial shield, core barrel, upper internal structure (UIS), and a primary vessel. Tests SM-4 and SM-5 are more complex models than SM-3. This paper presents comparisons of the ALICE-II code (Arbitrary Lagrangian Implicit-explicit Continuous Fluid Eulerian containment code - second version) with experiments SM-3 through SM-5. Two calculations are performed with ALICE-II on each of these three experiments, using both the pressure-time histories (p-t) and the pressure-volume relationships (p-v) as input to describe the energy source. Pressure profiles, dynamic strains, and vessel deformations are used as the basis of the comparison.

Ku, J.L.; Wang, C.Y.; Zeuch, W.R.

1983-01-01T23:59:59.000Z

334

Generic Assessment for Optimized Reactor Coolant System Hydrogen of a Four-loop Westinghouse Pressurized Water Reactor  

Science Conference Proceedings (OSTI)

The Chemistry, Fuel Reliability, and Material Reliability Programs at the Electric Power Research Institute (EPRI) have developed a comprehensive elevated reactor coolant system (RCS) hydrogen program that is focused on qualification of plant operation with dissolved hydrogen concentration in the RCS greater than 50 standard cubic centimeters per kilogram (scc/kg) (1.38 in.3/lbm), up to 60 scc/kg (1.66 in.3/lbm), to mitigate primary water stress corrosion cracking (PWSCC) in nickel-based alloys. Currentl...

2011-12-23T23:59:59.000Z

335

Materials Reliability Program: Loading Effects on the Low-Temperature Crack Propagation Phenomenon in 182 Weld Metal in a Pressurize d Water Reactor Environment (MRP-285)  

Science Conference Proceedings (OSTI)

This report summarizes results of a study of loading effects on the low-temperature crack propagation (LTCP) phenomenon in 182 weld metal in a pressurized water reactor (PWR) environment.

2010-12-20T23:59:59.000Z

336

Development of a New Flame Speed Vessel to Measure the Effect of Steam Dilution on Laminar Flame Speeds of Syngas Fuel Blends at Elevated Pressures and Temperatures  

E-Print Network (OSTI)

Synthetic gas, syngas, is a popular alternative fuel for the gas turbine industry, but the composition of syngas can contain different types and amounts of contaminants, such as carbon dioxide, methane, moisture, and nitrogen, depending on the industrial process involved in its manufacturing. The presence of steam in syngas blends is of particular interest from a thermo-chemical perspective as there is limited information available in the literature. This study investigates the effect of moisture content (0 ? 15% by volume), temperature (323 ? 423 K), and pressure (1 ? 10 atm) on syngas mixtures by measuring the laminar flame speed in a newly developed constant-volume, heated experimental facility. This heated vessel also broadens the experimental field of study in the authors? laboratory to low vapor pressure fuels and other vaporized liquids. The new facility is capable of performing flame speed experiments at an initial pressure as high as 30 atm and an initial temperature up to 600 K. Several validation experiments were performed to demonstrate the complete functionality of the flame speed facility. Additionally, a design-of-experiments methodology was used to study the mentioned syngas conditions that are relevant to the gas turbine industry. The design-of-experiments methodology provided the capability to identify the most influential factor on the laminar flame speed of the conditions studied. The experimental flame speed data are compared to the most up-to-date C4 mechanism developed through collaboration between Texas A&M and the National University of Ireland Galway. Along with good model agreement shown with all presented data, a rigorous uncertainty analysis of the flame speed has been performed showing an extensive range of values from 4.0 cm/s to 16.7 cm/s. The amount of carbon monoxide dilution in the fuel was shown to be the most influential factor on the laminar flame speed from fuel lean to fuel rich. This is verified by comparing the laminar flame speed of the atmospheric mixtures. Also, the measured Markstein lengths of the atmospheric mixtures are compared and do not demonstrate a strong impact from any one factor but the ratio of hydrogen and carbon monoxide plays a key role. Mixtures with high levels of CO appear to stabilize the flame structure of thermal-diffusive instability. The increase of steam dilution has only a small effect on the laminar flame speed of high-CO mixtures, while more hydrogen-dominated mixtures demonstrate a much larger and negative effect of increasing water content on the laminar flame speed.

Krejci, Michael

2012-05-01T23:59:59.000Z

337

Alternative Method for Performing Regulatory Guide 1.154 Pressurized Thermal Shock Analysis  

Science Conference Proceedings (OSTI)

Pressurized thermal shock (PTS) is a safety concern for some nuclear reactor pressure vessels with significant radiation embrittlement. This report presents a simplified method for assessing the failure risk associated with PTS and substantiating the benefit of actions taken to mitigate its effects.

1999-08-06T23:59:59.000Z

338

Alternative Method for Performing Regulatory Guide 1.154 Pressurized Thermal Shock Analysis  

Science Conference Proceedings (OSTI)

Pressurized thermal shock (PTS) is a safety concern for some nuclear reactor pressure vessels with significant radiation embrittlement. This report presents a simplified method for assessing the failure risk associated with PTS and substantiating the benefit of actions taken to mitigate its effects.

1997-04-02T23:59:59.000Z

339

Design for 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; Mieville, Rodney L.

1998-11-18T23:59:59.000Z

340

A Pressurized Water Reactor Plutonium Incinerator Based on Thorium Fuel and Seed-Blanket Assembly Geometry  

Science Conference Proceedings (OSTI)

A pressurized water reactor (PWR) fuel cycle is proposed, whose purpose is the elimination and degradation of weapons-grade plutonium. This Radkowsky thorium-fuel Pu incinerator (RTPI) cycle is based on a core and assemblies retrofittable to a Westinghouse-type PWR. The RTPI assembly, however, is a seed-blanket unit. The seed is supercritical, loaded with Pu-Zr alloy as fuel in a high moderator-to-fuel ratio configuration. The blanket is subcritical, loaded mainly with ThO{sub 2}, generating and burning {sup 233}U in situ. Blankets are loaded once every 6 yr. The seed fuel management scheme is based on three batches, with one-third of the seed modules replaced every year. The core generates 1100 MW(electric). Equilibrium conditions are achieved with the second seed loading. For equilibrium conditions, the annual average of disposed (loaded) Pu is 1210 kg, of which 702 kg are completely eliminated, and 508 kg are discharged, but with significantly degraded isotopics (i.e., with a high percentage of even mass isotopes). Spontaneous fissions per second in a gram of this degraded Pu are {approx}500, resulting in significantly increased proliferation resistance.Every 6 yr the blanket discharge contains 780 kg of {sup 233}U (including {sup 233}Pa) and 36 kg of {sup 235}U. However, the blankets are initially loaded with an amount of natural uranium selected such that these U fissile isotopes constitute only 12% of the total U discharge, a percentage equivalent to 20% {sup 235}U enrichment; hence, both the discharged uranium isotopics satisfy proliferation-resistant criteria.The RTPI control variables, namely, the moderator temperature coefficient, the reactivity per ppm boron, and the control rods worth, are about equal to those of a PWR. The RTPI spent-fuel stockpile ingestion toxicity over a period of ten million years is about the same as the counterpart toxicities of a regular, or a mixed-oxide (MOX), PWR. Compared with known PWR MOX variants, the RTPI is, per 1000 MW(electric) and per annum, a significantly more efficient incinerator of weapons-grade plutonium.

Galperin, A. [Ben-Gurion University of the Negev (Israel); Segev, M. [Ben-Gurion University of the Negev (Israel); Todosow, M. [Brookhaven National Laboratory (United States)

2000-11-15T23:59:59.000Z

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.


341

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.  

Science Conference Proceedings (OSTI)

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

Yun, D.; Taiwo, T. A.; Kim, T. K.; Mohamed, A.; Nuclear Engineering Division

2010-10-01T23:59:59.000Z

342

Nuclear reactor building  

DOE Patents (OSTI)

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05T23:59:59.000Z

343

Nuclear reactor building  

DOE Patents (OSTI)

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

344

High flux reactor  

DOE Patents (OSTI)

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Lake, James A. (Idaho Falls, ID); Heath, Russell L. (Idaho Falls, ID); Liebenthal, John L. (Idaho Falls, ID); DeBoisblanc, Deslonde R. (Summit, NJ); Leyse, Carl F. (Idaho Falls, ID); Parsons, Kent (Idaho Falls, ID); Ryskamp, John M. (Idaho Falls, ID); Wadkins, Robert P. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID); Fillmore, Gary N. (Idaho Falls, ID); Oh, Chang H. (Idaho Falls, ID)

1988-01-01T23:59:59.000Z

345

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

346

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

Science Conference Proceedings (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

347

Nuclear reactor overflow line  

DOE Patents (OSTI)

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

348

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

Science Conference Proceedings (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

349

Gas-phase chemistry during the conversion of cyclohexane to carbon: Flow reactor studies at low and intermediate pressure  

DOE Green Energy (OSTI)

The gas-phase branching during the conversion of cyclohexane to solid carbon has been measured in a high-temperature-flow reactor. The experiments show that cyclohexane decomposes into a broad distribution of hydrocarbons that further decompose into the more kinetically stable products hydrogen, methane, acetylene, ethylene, benzene, and PAH. At 1363 K, the evolution to these species occurs quickly. We also observe the buildup of significant amounts of aromatic molecules at later stages in the decomposition, with as much as 15% of the total carbon in PAH and 25% in benzene. At later stages, the gas-phase molecules react slowly, even though the system is not at equilibrium, because of their kinetic stability and the smaller radical pool. The decomposition does not appear to depend sensitively on pressure in the regime of 25 to 250 torr. Thus, to a first approximation, these results can be extrapolated to atmospheric pressure.

Osterheld, T.H.; Allendorf, M.D.; Larson, R.

1995-07-01T23:59:59.000Z

350

Program on Technology Innovation: Evaluation of Wear Characteristics of a Nanofluid in a Pressurized Water Reactor  

Science Conference Proceedings (OSTI)

A nanofluid is a fluid suspension of nanosized particles. It is a relatively new heat transfer fluid that is characterized by enhanced heat transfer properties over those of the base fluid. The enhanced heat transfer properties are desirable for potential use as the coolant in a nuclear reactor.

2008-01-31T23:59:59.000Z

351

Application of genetic algorithm to optimize burnable poison placement in pressurized water reactors  

Science Conference Proceedings (OSTI)

An efficient and a practical genetic algorithm tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem in the reference Three Mile Island-1 (TMI-1) core. Core BP optimization problem means developing a BP ... Keywords: burnable poison, decision variables, gadolinium, genetic algorithm, nuclear, optimization, reactor

Serkan Yilmaz; Kostadin Ivanov; Samuel Levine

2005-06-01T23:59:59.000Z

352

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Forsberg, Charles W. (Kingston, TN)

1987-01-01T23:59:59.000Z

353

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Forsberg, C.W.

1985-02-19T23:59:59.000Z

354

Real-Time 3D Simulation of a Pressurized Water Nuclear Reactor  

Science Conference Proceedings (OSTI)

Fuel assemblies are very expensive parts of the nuclear reactor. Initially they were used in Hungary for 3 years, now for 4 years and soon they will stay in the core for 5 years. Each year only 1/3rd, 1/4th later 1/5th of them is replaced, therefore ... Keywords: NPP simulation, parallel processing, real-time simulation, coupled thermo-hydraulics and neutron kinetics simulation

Janos Sebestyen Janosy; Andras Kereszturi; Gabor Hazi; Jozsef Pales; Endre Vegh

2010-03-01T23:59:59.000Z

355

LPG storage vessel cracking experience  

SciTech Connect

In order to evaluate liquefied petroleum gas (LPG) handling and storage hazards, Caltex Petroleum Corp. (Dallas) surveyed several installations for storage vessel cracking problems. Cracking was found in approximately one-third of the storage vessels. In most cases, the cracking appeared to be due to original fabrication problems and could be removed without compromising the pressure containment. Several in-service cracking problems found were due to exposure to wet hydrogen sulfide. Various procedures were tried in order to minimize the in-service cracking potential. One sphere was condemned because of extensive subsurface cracking. This article's recommendations concern minimizing cracking on new and existing LPG storage vessels.

Cantwell, J.E. (Caltex Petroleum Corp., P.O. Box 619500, Dallas, TX (US))

1988-10-01T23:59:59.000Z

356

LPG storage vessel cracking experience  

SciTech Connect

As part of an overall company program to evaluate LPG handling and storage hazards the authors surveyed several installations for storage vessel cracking problems. Cracking was found in approximately one third of the storage vessels. In most cases the cracking appeared due to original fabrication problems and could be removed without compromising the pressure containment. Several in-service cracking problems due to exposure to wet hydrogen sulfide were found. Various procedures were tried in order to minimize the in-service cracking potential. One sphere was condemned because of extensive subsurface cracking. Recommendations are made to minimize cracking on new and existing LPG storage vessels.

Cantwell, J.E.

1988-01-01T23:59:59.000Z

357

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

358

Calculation of releases of radioactive materials in gaseous and liquid effluents from pressurized water reactors (PWR-GALE Code). Revision 1  

SciTech Connect

This report revises the original issuance of NUREG-0017, ''Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE-Code)'' (April 1976), to incorporate more recent operating data now available as well as the results of a number of in-plant measurement programs at operating pressurized water reactors. The PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents (i.e., the gaseous and liquid source terms). The US Nuclear Regulatory Commission uses the PWR-GALE Code to determine conformance with the requirements of Appendix I to 10 CFR Part 50.

Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

1985-04-01T23:59:59.000Z

359

THE CNSG II--A CONCEPTUAL MERCHANT SHIP NUCLEAR REACTOR DESIGN  

SciTech Connect

The Consolidated Nuclear Steam Generator H consists of a pressurized water reactor, a steam generator, and a pressurizer combined in a sirgle pressure vessel. The design of the 66000 shaft horsepower system is presented, together with basic plant irformation for designs of 15000, 22000, and 30000 shaft horsepower. The economics, safety characteristics, and operational procedures of the plant are also discussed. (D.C.W.)

1963-09-01T23:59:59.000Z

360

Floating vessel  

SciTech Connect

The invention relates to a floating vessel which may be used in oil recovery. The assembly consists of a vertical column having a relatively small diameter. The column has a buoyancy capacity and is supplied with a ballast section having a larger diameter at its end. An upper structure is movably connected to the column. The column and the ballast chamber determine the limits of a shaft. The shaft is open at its lower end and is supplied with means to let fluid into the shaft over a relatively large area. (8 claims)

1974-05-14T23:59:59.000Z

Note: This page contains sample records for the topic "reactor pressure vessel" from the National Library of EnergyBeta (NLEBeta).
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361

Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features  

Science Conference Proceedings (OSTI)

A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

Winters, J. W.; Corletti, M. M.; Hayashi, Y.

2003-11-12T23:59:59.000Z

362

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

363

Digital computer code for simulating the dynamics of full-size dual-purpose desalting plants using a pressurized water reactor as a heat source  

SciTech Connect

A digital simulator was developed for use in calculating the dynamic response of full-size dual-purpose desalting plants. This simulator consists of a multistage flash (MSF) evaporator, a pressurized water reactor (PWR) as the heat source, a drumtype steam generator, and a turbine plant utilizing a back- pressure turbine. A bypass steam system was modeled to achieve flexible operation of the electric power and water portions of the combined plant. The proposed use of this simulator is to investigate various coupling and control schemes and to help determine possible problem areas in full back-pressure turbine dual-purpose desalting plant designs. (auth)

Delene, J.G.

1973-10-01T23:59:59.000Z

364

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM  

DOE Patents (OSTI)

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Moore, W.T.

1958-09-01T23:59:59.000Z

365

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955  

SciTech Connect

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied. Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)

1958-10-31T23:59:59.000Z

366

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD SEPTEMBER 9 TO OCTOBER 20, 1955  

SciTech Connect

Progress in the design, development, and construction of PWR power plant systems and components and PWR core and auxiliaries is summarized. The blanket assembly design is described and illustrated. Results of MTR evaluation of fuel element failure instrumentation are reported. Development of fabrication and testing tochaiques for clad fuel elements, fuel rods, plates, and assemblies is described. Investigations of fuel and cladding alloys include crystal structure and thermal stability determinations on U--Mo alloys, studies on the nature of the hydride phase formed during corrosion of gamma -phase alloys in high- temperature water, and specific heat, resistivity, and phase diagram studies of U- -Mo and U--Nb alloys. The equilibrium and kinetics in the system UO/sub 2/--O/ sub 2/ are being studied to gain information on the structure and stability of UO/ sub 2/ under various conditions. Results of irradiation tests on UO/sub 2/ samples and of thermal cycling tests of Zircaloy-2 clad UO/sub 2/ rods are reported. Corrosion test resuIts are summarized for unclad and Zircaloy-2 clad U- - Mo and U--Nb samples. The radiation induced volume change of prototype fuel reds has been investigated. Studies of the fabrication cladding, tensile properties, and corrosion of U-- Si systems are described. Corrosion tests are continuing on Zircaloy-2 clad U-- Zr fuel elements and on various experimental Al alloys for cladding applications. Work was continued on the preparation, corrosion and sinterability of pure UO/sub 2/ and UO/sub 2/ containing additives. Operation and chemical analysis of in-pile loop experiments are described. Results are reported from studies of the erosion of UO/sub 2/ in high-velocity coolant, decontamination of water by ion exchange resins, sorption of radioisotopes on stainless steel, and decontamination of corrosion loops. Work in reactor physics has included PWR control calculations using a 2-dimensional UNIVAC code, calculation of fission product activity in the primary coolant, and criticaiity studies on the Flexibie Critical Experiment and on a lattice of UO/ sub 2/ fuel reds in the TRX. Current PWR plant parameters are recapitulated. (D.E.B.)

1958-10-31T23:59:59.000Z

367

BWRVIP-244: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2010  

Science Conference Proceedings (OSTI)

This report provides 2010 results of the nondestructive evaluation NDE development task of the Boiling Water Reactor Vessel and Internals Project BWRVIP Inspection Focus Group. The scope of activity includes applications of various NDE techniques to boiling water reactor vessels and vessel internals components.

2010-12-23T23:59:59.000Z

368

A model for calculation of RCS pressure during reflux boiling under reduced inventory conditions and its assessment against PKL data. [Reactor Cooling Systems (RCS)  

Science Conference Proceedings (OSTI)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.

Palmrose, D.E. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Mandl, R.M. (Siemens AG, Berlin (Germany))

1991-01-01T23:59:59.000Z

369

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

370

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

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 is described in this disclosure. 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, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

371

Hydrostatic Pressure Retainment.  

E-Print Network (OSTI)

??There is a great deal of attention being concentrated on reducing the weight of pressure vessels and fuel/oxidizer tanks (tankage) by 10% to 20%. Most… (more)

Setlock, Robert J., Jr.

2004-01-01T23:59:59.000Z

372

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR  

DOE Patents (OSTI)

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14T23:59:59.000Z

373

SNAP II REACTOR CORE MATERIALS  

SciTech Connect

A survey was made to select the construction materials for the SDR-1 reactor core vessel and grid plates. Hastelloy C was selected for the reactor vessel, top grid plate, and bottom grid plate. Inconel X was selected for the core hold-down springs. (C.J.G.)

Facha, J.V.

1960-04-11T23:59:59.000Z

374

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

375

Fuel Cell Technologies Office: International Hydrogen Fuel and Pressure  

NLE Websites -- All DOE Office Websites (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

376

Program on Technology Innovation: Hybrid Models of Stress Corrosion Crack Propagation for Nickel Alloy Welds in Low-Electrochemical Potential (ECP) Pressurized Water Reactor (PWR) Primary Water Environments  

Science Conference Proceedings (OSTI)

EPRI has developed hybrid models of pressurized water reactor (PWR) primary water stress corrosion cracking (PWSCC) in nickel alloy welds.  These models are able to account for differences in tensile properties of each heat, applied stress intensity factor, dissolved hydrogen, water temperature, and the increase in local strain rate caused by the moving crack. The new models show promise for reducing uncertainty in predicting PWSCC for nickel alloy welds by a statistically and practically ...

2012-10-30T23:59:59.000Z

377

Steam Generator Management Program: Pressurized Water Reactor Generic Tube Degradation Predictions: Recirculating Steam Generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG Tubing  

Science Conference Proceedings (OSTI)

Mill-annealed Alloy 600 heat transfer tubing in pressurized water reactor (PWR) steam generators (SGs) has experienced numerous modes of degradation. This report describes predictive models for determining expected tube degradation in recirculating steam generators with Alloy 600TT, Alloy 690TT, and Alloy 800NG tubing. Predictions are based on operating experience with similar designs and use improvement factors to characterize benefits resulting from SG design and material ...

2013-12-17T23:59:59.000Z

378

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

Science Conference Proceedings (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

379

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

380

EDS V25 containment vessel explosive qualification test report.  

SciTech Connect

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

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

GAS COOLED NUCLEAR REACTOR STUDY. Final Report  

SciTech Connect

An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)

Thompson, A.S.

1956-07-31T23:59:59.000Z

382

PNL technical review of pressurized thermal-shock issues. [PWR  

SciTech Connect

Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur. Therefore, the near-term (less than one year) recommendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be developed by NRC for use during future licensee analyses. Finally, it was recommended that NRC upgrade nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirements.

Pedersen, L.T.; Apley, W.J.; Bian, S.H.; Defferding, L.J.; Morgenstern, M.H.; Pelto, P.J.; Simonen, E.P.; Simonen, F.A.; Stevens, D.L.; Taylor, T.T.

1982-07-01T23:59:59.000Z

383

MAAP5 BWR Vessel Penetration and Ex-Vessel Equipment Model Enhancement Description  

Science Conference Proceedings (OSTI)

This report describes proposed enhancements to the Modular Accident Analysis Program (MAAP) vessel penetration and ex-vessel equipment model for BWR designs. MAAP is an EPRI-owned and -licensed computer program that simulates the operation of light water and heavy water moderated nuclear power plants for both current and advanced light water reactor (ALWR) designs.The report explores the manner in which the in-core instrument tubes would respond during severe core damage events that ...

2013-02-25T23:59:59.000Z

384

Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.  

SciTech Connect

STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal resistance of a gas-filled gap.

Moisseytsev, A.; Sienicki, J. J.

2007-03-08T23:59:59.000Z

385

Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure (MRP-156)  

Science Conference Proceedings (OSTI)

The Industry Initiative on the Management of Materials Issues provides a proactive, safety-focused approach to the management of materials degradation. In support of this initiative, EPRI formed the Materials Degradation Assessment/Issue Management Table Ad-Hoc Committee and developed an Issue Management Table (IMT) for reactor coolant system components. This report provides initial input to the IMT to address the consequences of failure for the identified components in the reactor coolant system for ope...

2005-12-16T23:59:59.000Z

386

Prediction of Vessel Icing  

Science Conference Proceedings (OSTI)

Vessel icing from wave-generated spray is a severe hazard to expanded marine operations in high latitudes. Hardships in making observations during operations, combined with differences in vessel type and heading, have resulted in great ...

J. E. Overland; C. H. Pease; R. W. Preisendorfer; A. L. Comiskey

1986-12-01T23:59:59.000Z

387

Treating exhaust gas from a pressurized fluidized bed reaction system  

DOE Patents (OSTI)

Hot gases from a pressurized fluidized bed reactor system are purified. Under super atmospheric pressure conditions hot exhaust gases are passed through a particle separator, forming a filtrate cake on the surface of the separator, and a reducing agent--such as an NO{sub x} reducing agent (like ammonia)--is introduced into the exhaust gases just prior to or just after particle separation. The retention time of the introduced reducing agent is enhanced by providing a low gas velocity (e.g. about 1--20 cm/s) during passage of the gas through the filtrate cake while at super atmospheric pressure. Separation takes place within a distinct pressure vessel, the interior of which is at a pressure of about 2--100 bar, and introduction of reducing agent can take place at multiple locations (one associated with each filter element in the pressure vessel), or at one or more locations just prior to passage of clean gas out of the pressure vessel (typically passed to a turbine). 8 figs.

Isaksson, J.; Koskinen, J.

1995-08-22T23:59:59.000Z

388

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

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

389

TMI-2 Vessel Investigation Project integration report  

Science Conference Proceedings (OSTI)

The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-03-01T23:59:59.000Z

390

Vacuum Vessel Remote Handling  

E-Print Network (OSTI)

FIRE Vacuum Vessel and Remote Handling Overview B. Nelson, T. Burgess, T. Brown, H-M Fan, G. Jones #12;13 July 2002 Snowmass Review: FIRE Vacuum Vessel and Remote Handling 2 Presentation Outline · Remote Handling - Maintenance Approach & Component Classification - In-Vessel Transporter - Component

391

Dismantling techniques for reactor steel piping  

SciTech Connect

Two cutting techniques have been developed for dismantling the pipes connected to the Japan Power Demonstration Reactor (JPDR) pressure vessel. They are the rotary disk knife cutting system for dismantling relatively large pipes, such as the primary cooling system, and the shaped explosive cutting systems for cutting relatively small pipes in air or water. Basic cutting tests were performed to determine the optimum characteristics of the cutting systems and to conduct a safety evaluation by studying the effects of blasting on surrounding areas. Mock-up tests confirmed the applicability of the newly developed dismantling systems for JPDR dismantlement by successfully cutting test pipes with these systems.

Yanagihara, S.; Hiraga, F.; Nakamura, H. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

1989-08-01T23:59:59.000Z

392

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

393

Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980  

SciTech Connect

Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

Not Available

1980-05-01T23:59:59.000Z

394

PR-EDB: Power Reactor Embrittlement Database - Version 3  

Science Conference Proceedings (OSTI)

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for backend data storage, and Microsoft Excel for plotting graphs. This software package is compatible with Windows (98 or higher) and has been built with a highly versatile user interface. PR-EDB Version 3.0 also contains an "Evaluated Residual File" utility for generating the evaluated processed files used for radiation embrittlement study.

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

2008-03-01T23:59:59.000Z

395

Thermal Expansion Coefficient of Steels Used in LWR Vessels  

Science Conference Proceedings (OSTI)

Because of the impact that melt relocation and vessel failure have on subsequent progression and associated consequences of a Light Water Reactor (LWR) accident, it is important to accurately predict the heat-up and relocation of materials within the reactor vessel and heat transfer to and from the reactor vessel. Accurate predictions of such heat transfer phenomena require high temperature thermal properties. However, a review of vessel and structural steel material properties in severe accident analysis codes reveals that the required high temperature material properties are extrapolated with little, if any, data above 700ºC. To reduce uncertainties in predictions relying upon this extrapolated high temperature data, new thermal expansion data were obtained using pushrod dilatometry techniques for two metals used in LWR vessels: SA 533 Grade B, Class 1 (SA533B1) low alloy steel, which is used to fabricate most US LWR reactor vessels; and Type 304 Stainless Steel (SS304), which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data and compares it to existing, lower temperature data in the literature.

Joshua E. Daw; Joy L. Rempe; Darrell L. Knudson; John C. Crepeau

2008-05-01T23:59:59.000Z

396

High temperature thermal properties for metals used in LWR vessels  

Science Conference Proceedings (OSTI)

Because of the impact that melt relocation and vessel failure has on subsequent progression and associated consequences of an Light Water Reactor (LWR) accident, it is important to accurately predict the heatup and relocation of materials within the reactor vessel and heat transfer to and from the reactor vessel. Accurate predictions of such heat transfer phenomena require high temperature thermal properties. However, a review of vessel and structural steel material properties in severe accident analysis codes reveals that the required high temperature material properties are extrapolated, with little if any, data above 700 ºC. To reduce uncertainties in predictions relying upon this extrapolated high temperature data, INL obtained data using laser-flash thermal diffusivity techniques for two metals used in LWR vessels: SA533B1 carbon steel, which is used to fabricate most US LWR reactor vessels; and SS304, which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data, compares it to existing data in the literature, and provides recommended correlations for thermal properties based on this data.

Joy L. Rempe

2008-01-01T23:59:59.000Z

397

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

398

AN EVALUATION OF HEAVY WATER REACTORS FOR POWER  

DOE Green Energy (OSTI)

Reference designs for pressurized and direct-boiling D/sub 2/O reactors were prepared for electrical outputs of 20, 100, and 250 electrical Mw. A number of possible core designs were considered and those utilized which seemed most appropriate to give low-cost power. The technology and costs available today were employed in the preparation of the over-all plant designs. The Consolidated Western Steel Division of U. S. Steel Corporation assisted by preparing a comprehensive report on the design of large pressure vessels and containment vessels. Zr-clad U fuel elements were used as the study basis, but the effect of using UO/sub 2/ and stainless steel cladding was also considered. The principal results found were: (1) Over a wide range of operating conditions snd economic situations, enriched U (up to perhaps 1.4% U/sup 235/) is presently more economic to employ in D/sub 2/O reactors than is natural U. (2) In the longer range, the use of natural U may become more economic as Zr fabrication costs decrease, continuous charge-discharge devices are developed to permit longer exposure levels, and pressure-vessel technology advances so that the large critical masses and core diameters required are not such sn economic penalty on the natural U. The results agree quite well with the data and discussions of the Canadians. (auth)

Herron, D.P.; Newkirk, W.H.; Puishes, A.

1957-10-01T23:59:59.000Z

399

Analysis of hydrogen vehicles with cryogenic high pressure storage  

DOE Green Energy (OSTI)

Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LIQ) or ambient-temperature compressed hydrogen (CH2). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

Aceves, S. M.; Berry, G. D.

1998-06-19T23:59:59.000Z

400

BWRVIP-269: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2012  

Science Conference Proceedings (OSTI)

The purpose of this report is to describe the results of nondestructive evaluation (NDE) activities conducted in the previous year within the NDE Development task of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Inspection Focus Group. The scope of the ongoing NDE Development task includes the reactor vessel and its internal components. This task attempts to develop solutions for the more difficult ...

2012-12-12T23:59:59.000Z

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

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME I. SUMMARY REPORT  

SciTech Connect

Capital investments and the cost of power were estimated for 21 heavy- water-moderated, natural-uraniumfueled power-reactor plants, ranging in capacity from 100 to 460 Mw(e). Comparisons were made of hot- and coldmoderator reactors and of the relative merits of pressuretube and pressure-vessel designs. Reactors cooled with liquid D/sub 2/O, boiling D/sub 2/O, /sub 2/O steam, and helium were evalunted. A cold-moderator pressure-tube reactor cooled with boiling D/sub 2/O shows the most economic promise of the D/sub 2/Omoderated reactor systems studied to date. Reactors of this type have sufficient reactivity to permit satisfactory fuel exposures, but the development of additional technology is a prerequisite for optimum designs. At capacities of 300 and 400 Mw(e), the estimated power costs from the current designs of boiling-D/sub 2/O pressure-tabe reactor plants are 11.3 and 9.8 mills/kwh, respectively. From liquid-D/sub 2/-cooled concepts of comparable capacities the indicated power costs are 7 to 20% higher. With an active development program, a power cost of 8.0 to 8.5 mills/kwh may be attained in a 300 Mw(e) boiling-D/sub 2/O reactor plant within the next decade. (auth)

1960-06-01T23:59:59.000Z

402

Systems Engineering of Chemical Hydride, Pressure Vessel ...  

BoP Equipment Equations/Assumptions: ... Material and Synthetic Process Cost for raw material (precursor) and estimate for processing ($/g)

403

X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor  

SciTech Connect

Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented.

Magelssen, G.R.

1979-01-01T23:59:59.000Z

404

Nuclear reactor melt-retention structure to mitigate direct containment heating  

DOE Patents (OSTI)

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Tutu, Narinder K. (Manorville, NY); Ginsberg, Theodore (East Setauket, NY); Klages, John R. (Mattituck, NY)

1991-01-01T23:59:59.000Z

405

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

406

Coal gasification vessel. [Patent application  

DOE Patents (OSTI)

A vessel system comprises an outer shell of carbon fibers held in a binder, a coolant circulation mechanism and control mechanism and an inner shell 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 can be computer controlled and can be used to monitor and modulate the coolant which is provided through the circulation mechanism for cooling and protecting the carbon fiber and outer shell. The control mechanism is also used to locate any isolated hot spots which may occur through the local disintegration of the inner refractory shell.

Loo, B.W.

1981-03-17T23:59:59.000Z

407

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

Science Conference Proceedings (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

408

Stationary low power reactor No. 1 (SL-1) accident site decontamination & dismantlement project  

SciTech Connect

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft{sup 2} two story, cinder block administrative building; two 4,000 ft{sup 2} single story, steel frame office buildings; a 850 ft{sup 2} steel framed, metal sided PL condenser building, and a 550 ft{sup 2} steel framed decontamination and laydown building.

Perry, E.F.

1995-11-01T23:59:59.000Z

409

Nuclear Fuel Cycle Cost Comparison Between Once-Through and Plutonium Single-Recycling in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Within the context of long-term waste management and sustainable nuclear fuel supply, there continue to be discussions regarding whether the United States should consider recycling of light-water reactor (LWR) spent nuclear fuel (SNF) for the current fleet of U.S. LWRs. This report presents a parametric study of equilibrium fuel cycle costs for an open fuel cycle without plutonium recycling (once-through) and with plutonium recycling (single-recycling using mixed-oxide, or MOX, fuel), assuming an all-pre...

2009-02-25T23:59:59.000Z

410

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

King, L.D.P.

1959-09-01T23:59:59.000Z

411

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

412

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

413

Instrumentation of a prestressed concrete containment vessel model  

Science Conference Proceedings (OSTI)

A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a {open_quotes}Tensmeg{close_quotes}{reg_sign} gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed.

Hessheimer, M.F.; Rightley, M.J. [Sandia National Labs., Albuquerque, NM (United States); Matsumoto, T. [Nuclear Power Engineering Corp., Tokyo (Japan)] [and others

1995-09-01T23:59:59.000Z

414

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME II. ENGINEERING STUDIES AND TECHNICAL DATA  

SciTech Connect

The results of preliminary design and evaluation studies of various concepts of a power reactor that is moderated by heavy water and fueled with natural uranium are presented. Twenty-nine conceptal designs were developed for reactors ranging in capacity from 100 Mwe to 460 Mwe. Resigns were prepared for hot- and cold-moderator reac tors of the pressure vessel type, with liquid D/sub 2/O, boiling D/sub 2/O, E/sub 2/O steam, and helium as coolants. Also studied were cold-moderator pressure tube reactors cooled with liquid D/sub 2/O and boiling D/sub 2/O. The repont includes the results of engineering studies of the reactor systems, electrical generation facilities, and auxiliary equipment. (auth)

1960-12-01T23:59:59.000Z

415

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

416

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

417

Design and Manufacture of the Storage Cask for the Old Reactor Internals  

SciTech Connect

Mitsubishi Heavy Industries, Ltd. (MHI) completed replacement work of upper reactor internals (UCI) and lower reactor internals (LCI) of the pressurized water reactor in Shikoku Electric Power Company's Ikata Unit No.1 by 'the all-in-one-piece extraction method' introduced in the document of [ICONE14-89233]. In the pressurized water reactor (PWR) plant, the UCI are usually removed from the reactor vessel (RV) independently and reinstalled into the RV again every refueling outage. The LCI are independently able to be removed from the RV and reinstalled again during in-service inspection, too. In the boiling water reactor (BWR) plant, there are several cases of replacing BWR shrouds by cutting small and containing in a container. But no replacement of all reactor internals (CI) for the PWR, in one piece without splitting or cutting, has been reported. The purpose of this paper is to introduce the key points about the design and manufacture of the storage cask for old reactor internals in the replacement work by 'the all-in-one-piece extraction method'. (author)

Yasuhiro Tomiita [Mitsubishi Heavy Industries, Ltd. (Japan)

2006-07-01T23:59:59.000Z

418

Cover Heated, Open Vessels  

SciTech Connect

This revised ITP steam tip sheet on covering heated, open vessels provides how-to advice for improving industrial steam systems using low-cost, proven practices and technologies.

2006-01-01T23:59:59.000Z

419

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

420

Reconnecting broken blood vessels  

NLE Websites -- All DOE Office Websites (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.

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421

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, moving particle semi implicit.

Suprijadi; Mohamad Rendi; Petrus Subekti; Sparisoma Viridi

2013-12-14T23:59:59.000Z

422

High temperature post-irradiation performance of spent pressurized-water-reactor fuel rods under dry-storage conditions  

Science Conference Proceedings (OSTI)

Post-irradiation studies on failure mechanisms of well characterized PWR rods were conducted for up to a year at 482, 510 and 571/sup 0/C in unlimited air and inert gas atmospheres. No cladding breaches occurred even though the tests operated many orders of magnitude longer in time than the lifetime predicted by Blackburn's analyses. The extended lifetime is due to significant creep strain of the Zircaloy cladding which decreases the internal rod pressures. The cladding creep also contributes to radial cracks, through the external oxide and internal FCCI layers, which propagated into and arrested in an oxygen stabilized ..cap alpha..-Zircaloy layer. There were no signs of either additional cladding hydriding, stress-corrosion cracking or fuel pellet degradation. Using the Larson-Miller formulization, a conservative maximum storage temperature of 400/sup 0/C is recommended to ensure a 1000-year cladding lifetime. This accounts for crack propagation and assumes annealing of the irradiation-hardened cladding.

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

1981-06-01T23:59:59.000Z

423

BWRVIP-34-A: BWR Vessel and Internals Project, Technical Basis for Part Circumference Weld Overlay Repair of Vessel Internal Core Sp ray Piping  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This report summarizes the results of the design and analysis activities and the testing programs conducted to provide BWR utilities with a contingency repair option for internal core spray piping for BWR2/6 plants. A previous version of this report was published as BWRVIP-34 (TR-108198). This report (BWRVIP-34...

2008-03-11T23:59:59.000Z

424

Electrochemistry of Water-Cooled Nuclear Reactors  

DOE Green Energy (OSTI)

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy, Amit Jain, Han Sang Kim, Vishisht Gupta; Jonathan Pitt

2006-08-08T23:59:59.000Z

425

Optical Measurement Technologies for High Temperature, Radiation Exposure, and Corrosive Environments—Significant Activities and Findings: In-vessel Optical Measurements for Advanced SMRs  

Science Conference Proceedings (OSTI)

Development of advanced Small Modular Reactors (aSMRs) is key to providing the United States with a sustainable, economically viable, and carbon-neutral energy source. The aSMR designs have attractive economic factors that should compensate for the economies of scale that have driven development of large commercial nuclear power plants to date. For example, aSMRs can be manufactured at reduced capital costs in a factory and potentially shorter lead times and then be shipped to a site to provide power away from large grid systems. The integral, self-contained nature of aSMR designs is fundamentally different than conventional reactor designs. Future aSMR deployment will require new instrumentation and control (I&C) architectures to accommodate the integral design and withstand the extreme in-vessel environmental conditions. Operators will depend on sophisticated sensing and machine vision technologies that provide efficient human-machine interface for in-vessel telepresence, telerobotic control, and remote process operations. The future viability of aSMRs is dependent on understanding and overcoming the significant technical challenges involving in-vessel reactor sensing and monitoring under extreme temperatures, pressures, corrosive environments, and radiation fluxes

Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong (Amy); Suter, Jonathan D.

2012-09-01T23:59:59.000Z

426

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

427

CC Pressure Test  

SciTech Connect

The inner vessel heads including bypass and beam tubes had just been welded into place and dye penetrant checked. The vacuum heads were not on at this time but the vacuum shell was on covering the piping penetrating into the inner vessel. Signal boxes with all feed through boards, the instrumentation box, and high voltage boxes were all installed with their pump outs capped. All 1/4-inch instrumentation lines were terminated at their respective shutoff valves. All vacuum piping used for pumping down the inner vessel was isolated using o-ring sealed blind flanges. PV215A (VAT Series 12), the 4-inch VRC gate valve isolating the cyropump, and the rupture disk had to be removed and replaced with blind flanges before pressurizing due to their pressure limitations. Stresses in plates used as blind flanges were checked using Code calcualtions. Before the CC test, vacuum style blanks and clamps were hydrostatically pressure tested to