National Library of Energy BETA

Sample records for reactor pressure vessel

  1. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

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

    1994-10-04

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

  2. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

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

  3. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, James K.

    1994-01-11

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

  4. Reactor pressure vessel vented head

    DOE Patents [OSTI]

    Sawabe, J.K.

    1994-01-11

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

  5. Reactor pressure vessel with forged nozzles

    DOE Patents [OSTI]

    Desai, Dilip R.

    1993-01-01

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

  6. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect (OSTI)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  7. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel...

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

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

  8. Nuclear reactor pressure vessel support system

    DOE Patents [OSTI]

    Sepelak, George R.

    1978-01-01

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

  9. Neutron shielding panels for reactor pressure vessels

    DOE Patents [OSTI]

    Singleton, Norman R.

    2011-11-22

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

  10. Midland reactor pressure vessel flaw distribution

    SciTech Connect (OSTI)

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

    1993-12-01

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

  11. Reactor pressure vessel structural integrity research

    SciTech Connect (OSTI)

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

    1995-04-01

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

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

    SciTech Connect (OSTI)

    Wang, Jy-An John

    2010-08-01

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

  13. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials

    Office of Energy Efficiency and Renewable Energy (EERE)

    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...

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

    SciTech Connect (OSTI)

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

    1994-10-01

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

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

    SciTech Connect (OSTI)

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

    2015-07-01

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

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

    SciTech Connect (OSTI)

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

    2011-01-01

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

  17. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect (OSTI)

    Wang, Jy-An John

    2015-08-01

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

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

    SciTech Connect (OSTI)

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

    2014-10-28

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

  19. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges

    Broader source: Energy.gov [DOE]

    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...

  20. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    SciTech Connect (OSTI)

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

    2015-03-01

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

  1. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    SciTech Connect (OSTI)

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

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

    SciTech Connect (OSTI)

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

    2012-01-01

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

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

    SciTech Connect (OSTI)

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

    2006-07-01

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

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

    SciTech Connect (OSTI)

    H. D. Gougar; C. B. Davis

    2006-04-01

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

  5. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment

    Broader source: Energy.gov [DOE]

    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...

  6. Update to Risk-Informed Pressurized Water Reactor Vessel 10 to 20 Year Inspection Interval Extension

    SciTech Connect (OSTI)

    Palm, Nathan A.; Bishop, Bruce A.; Boggess, Cheryl L.

    2006-07-01

    The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper. (authors)

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

    SciTech Connect (OSTI)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

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

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

    SciTech Connect (OSTI)

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

    2006-02-15

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

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

    DOE Patents [OSTI]

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

    1999-07-27

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

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

    DOE Patents [OSTI]

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

    1999-01-01

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

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

    SciTech Connect (OSTI)

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

    2007-03-21

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

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01

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

  13. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

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

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

    SciTech Connect (OSTI)

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

    2015-10-01

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

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

    SciTech Connect (OSTI)

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

    2012-09-01

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

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

    SciTech Connect (OSTI)

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

    1998-03-01

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

  17. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect (OSTI)

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  18. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

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

  19. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

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

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

    SciTech Connect (OSTI)

    David H. Bothell

    2000-04-30

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

  1. Dual shell pressure balanced vessel

    DOE Patents [OSTI]

    Fassbender, Alexander G.

    1992-01-01

    A dual-wall pressure balanced vessel for processing high viscosity slurries at high temperatures and pressures having an outer pressure vessel and an inner vessel with an annular space between the vessels pressurized at a pressure slightly less than or equivalent to the pressure within the inner vessel.

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01

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

  3. HFIR (High Flux Isotope Reactor) pressure vessel and structural components materials surveillance program: Supplement 1

    SciTech Connect (OSTI)

    Cheverton, R.D.; McGinty, D.M.; McWherter, J.R.; Nanstad, R.K.

    1987-10-01

    Extending the life of the HFIR vessel by the proposed 10 effective full-power years is contingent upon a continuation of the materials surveillance program and the application of hydrostatic proof testing. As a part of the surveillance program, Charpy V-notch (CVN) specimens of shell, weld and nozzle materials are installed adjacent to the inner surface of the vessel and are removed periodically for testing to determine the radiation-induced increase in the nil-ductility transition temperature. Hydro testing is conducted to prove that a critical combination of flaw size, stress and fracture toughness does not exist. Information from the materials surveillance program is used in a fracture mechanics analysis to confirm that the hydro-test pressure being applied is appropriate for the desired life extension of the vessel. This report specifies (1) the number, type, location and schedule for removal-testing of the CVN specimens for the continuing materials surveillance program, and (2) the procedures and test conditions for the hydro test.

  4. Sapphire tube pressure vessel

    DOE Patents [OSTI]

    Outwater, John O. (Cambridge, MA)

    2000-01-01

    A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

  5. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    SciTech Connect (OSTI)

    Ren, Weiju; Terry, Totemeier

    2006-10-01

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

  6. GOLD PRESSURE VESSEL SEAL

    DOE Patents [OSTI]

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  7. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

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

    1980-05-09

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

  8. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E.

    1987-01-01

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

  9. High pressure storage vessel

    DOE Patents [OSTI]

    Liu, Qiang

    2013-08-27

    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.

  10. Code System to Calculate Pressure Vessel Failure Probabilities.

    Energy Science and Technology Software Center (OSTI)

    2001-03-27

    Version 00 OCTAVIA (Operationally Caused Transients And Vessel Integrity Analysis) calculates the probability of pressure vessel failure from operationally-caused pressure transients which can occur in a pressurized water reactor (PWR). For specified vessel and operating environment characteristics the program computes the failure pressure at which the vessel will fail for different-sized flaws existing in the beltline and the probability of vessel failure per reactor year due to the flaw. The probabilities are summed over themore » various flaw sizes to obtain the total vessel failure probability. Sensitivity studies can be performed to investigate different vessel or operating characteristics in the same computer run.« less

  11. Reactor vessel seal service fixture

    DOE Patents [OSTI]

    Ritz, W.C.

    1975-12-01

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

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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...

  13. Reactor vessel annealing system

    DOE Patents [OSTI]

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

    1991-01-01

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

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

    SciTech Connect (OSTI)

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7 × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Edmondson, Philip D.; Miller, Michael K.; Powers, Kathy A.; Nanstad, Randy K.

    2015-12-29

    Surveillance samples of a low copper (nominally 0.05 wt.% Cu) forging and a higher copper (0.23 wt.% Cu) submerged arc weld from the R. E. Ginna reactor pressure vessel have been characterized by atom probe tomography (APT) after exposure to three levels of neutron irradiation, i.e., fluences of 1.7, 3.6 and 5.8 × 1023 n.m–2 (E > 1 MeV), and inlet temperatures of ~289 °C (~552 °F). As no copper-enriched precipitates were observed in the low copper forging, and the measured copper content in the ferrite matrix was 0.04± <0.01 at.% Cu, after neutron irradiation to a fluence of 1.7more » × 1023 n.m–3, this copper level was below the solubility limit. A number density of 2 × 1022 m–3 of Ni–, Mn– Si-enriched precipitates with an equivalent radius of gyration of 1.7 ± 0.4 nm were detected in the sample. However, Cu-, Ni-, Mn-enriched precipitates were observed in specimens cut from different surveillance specimens from the same forging material in which the overall measured copper level was 0.08± <0.01 at.% (fluence of 3.6 × 1023 n.m–3) and 0.09± <0.01 at.% Cu (fluence of 5.8 × 1023 n.m–3). Therefore, these slightly higher copper contents were above the solubility limit of Cu under these irradiation conditions. A best fit of all the composition data indicated that the size and number density of the Cu-enriched precipitates increased slightly in both size and number density by additional exposure to neutron irradiation. High number densities of Cu-enriched precipitates were observed in the higher Cu submerged arc weld for all irradiated conditions. The size and number density of the precipitates in the welds were higher than in the same fluence forgings. Some Cu-enriched precipitates were found to have Ni-, Mn- Si-, and P-enriched regions on their surfaces suggesting a preferential nucleation site. Furthermore, atom maps revealed P, Ni, and Mn segregation to, and preferential precipitation of, Cu-enriched precipitates over the surface of a grain

  16. Cavity closure arrangement for high pressure vessels

    DOE Patents [OSTI]

    Amtmann, Hans H.

    1981-01-01

    A closure arrangement for a pressure vessel such as the pressure vessel of a high temperature gas-cooled reactor wherein a liner is disposed within a cavity penetration in the reactor vessel and defines an access opening therein. A closure is adapted for sealing relation with an annular mounting flange formed on the penetration liner and has a plurality of radially movable locking blocks thereon having outer serrations adapted for releasable interlocking engagement with serrations formed internally of the upper end of the penetration liner so as to effect high strength closure hold-down. In one embodiment, ramping surfaces are formed on the locking block serrations to bias the closure into sealed relation with the mounting flange when the locking blocks are actuated to locking positions.

  17. Level indicator for pressure vessels

    DOE Patents [OSTI]

    Not Available

    1982-04-28

    A liquid-level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic-field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal-processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  18. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

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

    2013-03-19

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

  20. Pressurized water reactor flow skirt apparatus

    DOE Patents [OSTI]

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

    2016-04-05

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

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

    SciTech Connect (OSTI)

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

    1998-03-01

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

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

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger

    2013-07-01

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

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

    SciTech Connect (OSTI)

    Sullivan, Edmund J.; Anderson, Michael T.

    2014-06-10

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

  4. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

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

    1998-08-25

    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.

  5. Lightweight bladder lined pressure vessels

    DOE Patents [OSTI]

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    1998-01-01

    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.

  6. Investigation of vessel exterior air cooling for an HLMC reactor

    SciTech Connect (OSTI)

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

    2000-07-01

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

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

    SciTech Connect (OSTI)

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

    2000-01-13

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

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

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

    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

  9. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect (OSTI)

    BOWMAN, B.R.

    2002-10-23

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

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

    DOE Patents [OSTI]

    Isaksson, J.M.

    1996-06-18

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

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

    DOE Patents [OSTI]

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01

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

  12. PURE NIOBIUM AS A PRESSURE VESSEL MATERIAL

    SciTech Connect (OSTI)

    Peterson, T. J.; Carter, H. F.; Foley, M. H.; Klebaner, A. L.; Nicol, T. H.; Page, T. M.; Theilacker, J. C.; Wands, R. H.; Wong-Squires, M. L.; Wu, G.

    2010-04-09

    Physics laboratories around the world are developing niobium superconducting radio frequency (SRF) cavities for use in particle accelerators. These SRF cavities are typically cooled to low temperatures by direct contact with a liquid helium bath, resulting in at least part of the helium container being made from pure niobium. In the U.S., the Code of Federal Regulations allows national laboratories to follow national consensus pressure vessel rules or use of alternative rules which provide a level of safety greater than or equal to that afforded by ASME Boiler and Pressure Vessel Code. Thus, while used for its superconducting properties, niobium ends up also being treated as a material for pressure vessels. This report summarizes what we have learned about the use of niobium as a pressure vessel material, with a focus on issues for compliance with pressure vessel codes. We present results of a literature search for mechanical properties and tests results, as well as a review of ASME pressure vessel code requirements and issues.

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

    SciTech Connect (OSTI)

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

    2012-09-17

    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

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

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

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

  15. Investigation of Cracked Lithium Hydride Reactor Vessels

    SciTech Connect (OSTI)

    bird, e.l.; mustaleski, t.m.

    1999-06-01

    Visual examination of lithium hydride reactor vessels revealed cracks that were adjacent to welds, most of which were circumferentially located in the bottom portion of the vessels. Sections were cut from the vessels containing these cracks and examined by use of the metallograph, scanning electron microscope, and microprobe to determine the cause of cracking. Most of the cracks originated on the outer surface just outside the weld fusion line in the base material and propagated along grain boundaries. Crack depths of those examined sections ranged from {approximately}300 to 500 {micro}m. Other cracks were reported to have reached a maximum depth of 1/8 in. The primary cause of cracking was the creation of high tensile stresses associated with the differences in the coefficients of thermal expansion between the filler metal and the base metal during operation of the vessel in a thermally cyclic environment. This failure mechanism could be described as creep-type fatigue, whereby crack propagation may have been aided by the presence of brittle chromium carbides along the grain boundaries, which indicates a slightly sensitized microstructure.

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

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01

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

  17. Design Considerations For Blast Loads In Pressure Vessels.

    SciTech Connect (OSTI)

    Rodriguez, E. A.; Nickell, Robert E.; Pepin, J. E.

    2007-01-01

    Los Alamos National Laboratory (LANL), under the auspices of the U.S. Department of Energy (DOE) and the National Nuclear Security Administration (NNSA), conducts confined detonation experiments utilizing large, spherical, steel pressure vessels to contain the reaction products and hazardous materials from high-explosive (HE) events. Structural design and analysis considerations include: (a) Blast loading phase (i.e., impulsive loading); (b) Dynamic structural response; (c) Fragment (i.e., shrapnel) generation and penetration; (d) Ductile and non-ductile fracture; and (e) Design Criteria to ASME Code Sec. VIII, Div. 3, Impulsively Loaded Vessels. These vessels are designed for one-time-use only, efficiently utilizing the significant plastic energy absorption capability of ductile vessel materials. Alternatively, vessels may be designed for multiple-detonation events, in which case the material response is restricted to elastic or near-elastic range. Code of Federal Regulations, Title 10 Part 50 provides requirements for commercial nuclear reactor licensing; specifically dealing with accidental combustible gases in containment structures that might cause extreme loadings. The design philosophy contained herein may be applied to extreme loading events postulated to occur in nuclear reactor and non-nuclear systems or containments.

  18. Conformable pressure vessel for high pressure gas storage

    DOE Patents [OSTI]

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  19. International Hydrogen Fuel and Pressure Vessel Forum | Department of

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

    Energy Hydrogen Fuel and Pressure Vessel Forum International Hydrogen Fuel and Pressure Vessel Forum The U.S. Department of Energy (DOE) and Tsinghua University in Beijing co-hosted the International Hydrogen Fuel and Pressure Vessel Forum on September 27-29, 2010 in Beijing, China. High pressure vessel experts gathered to share lessons learned from compressed natural gas (CNG) and hydrogen vehicle deployments, and to identify R&D needs to aid the global harmonization of regulations,

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  1. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment- Volume 1

    Broader source: Energy.gov [DOE]

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment - Volume 1 August 2013

  2. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment- Volume 2

    Office of Energy Efficiency and Renewable Energy (EERE)

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment - Volume 2 August 2013

  3. International Hydrogen Fuel and Pressure Vessel Forum - Presentations |

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

    Department of Energy International Hydrogen Fuel and Pressure Vessel Forum - Presentations International Hydrogen Fuel and Pressure Vessel Forum - Presentations These presentations were given at the International Hydrogen Fuel and Pressure Vessel Forum held September 27-29, 2010 in Beijing, China. September 27, 2010 Keynote: Status and Progress in Research, Development and Demonstration of Hydrogen-Compressed Natural Gas Vehicles in China Professor Z.Q. Mao Tsinghua University and Chair of

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

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

    Department of Energy Fuel and Pressure Vessel Forum 2010 Proceedings International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings Proceedings from the forum, which took place in Beijing, China, on September 27-29, 2010. International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings (284.25 KB) More Documents & Publications Workshop Notes from ""Compressed Natural Gas and Hydrogen Fuels: Lessons Learned for the Safe Deployment of Vehicles"" Workshop,

  5. High-pressure Storage Vessels for Hydrogen, Natural Gas and

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

    Hydrogen-Natural Gas Blends | Department of Energy High-pressure Storage Vessels for Hydrogen, Natural Gas and Hydrogen-Natural Gas Blends High-pressure Storage Vessels for Hydrogen, Natural Gas and Hydrogen-Natural Gas Blends These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpv_lynch.pdf (4.21 MB) More Documents & Publications Properties, Behavior and Material Compatibility of Hydrogen, Natural Gas

  6. Stress analysis and evaluation of a rectangular pressure vessel...

    Office of Scientific and Technical Information (OSTI)

    States)) 42 ENGINEERING; 12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; PRESSURE VESSELS; STRESS ANALYSIS; RADIOACTIVE WASTE STORAGE;...

  7. Reactor vessel using metal oxide ceramic membranes

    DOE Patents [OSTI]

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

    1992-08-11

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

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

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

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

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

    DOE Patents [OSTI]

    Isaksson, J.

    1996-02-20

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

  10. International Hydrogen Fuel and Pressure Vessel Forum 2010 Proceedings

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

    International Hydrogen Fuel and Pressure Vessel Forum 2010 Beijing, China September 27-29, 2010 Background The China Association for Hydrogen Energy, the Engineering Research Center of High Pressure Process Equipment and Safety of the Ministry of Education in China, and the United States Department of Energy (DOE) conducted the International Hydrogen Fuel and Pressure Vessel Forum 2010, at Tsinghua University in Beijing, China, September 27-29, 2010. The Forum was co-organized by Professor Z.Q.

  11. An optimization study for the reactor vessel auxiliary cooling system of a pool liquid-metal reactor

    SciTech Connect (OSTI)

    Tzanos, C.P.; Tessier, H.; Pedersen, D.R. )

    1991-04-01

    This paper reports on the effects of design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) of a pool liquid-metal reactor (LMR). These parameters include stack height, size of the airflow gap, system pressure loss, fins on the guard vessel or the baffle wall, and repeated ribs on the airflow channel walls. As a measure of performance , the peak sodium pool temperature during transient following a reactor scram from full power was used. Horizontal ribs with a 0.003-m height and a 0.015-m pitch gave the best performance, i.e., the lowest peak sodium pool temperature during the scram transient. For a 3500-MW(thermal) LMR, they gave peak hot pool and peak cladding temperatures that were 52{degrees}C lower than those obtained with a reference RVACS having smooth airflow channel walls.

  12. Lightweight cryogenic-compatible pressure vessels for vehicular fuel storage

    DOE Patents [OSTI]

    Aceves, Salvador; Berry, Gene; Weisberg, Andrew H.

    2004-03-23

    A lightweight, cryogenic-compatible pressure vessel for flexibly storing cryogenic liquid fuels or compressed gas fuels at cryogenic or ambient temperatures. The pressure vessel has an inner pressure container enclosing a fuel storage volume, an outer container surrounding the inner pressure container to form an evacuated space therebetween, and a thermal insulator surrounding the inner pressure container in the evacuated space to inhibit heat transfer. Additionally, vacuum loss from fuel permeation is substantially inhibited in the evacuated space by, for example, lining the container liner with a layer of fuel-impermeable material, capturing the permeated fuel in the evacuated space, or purging the permeated fuel from the evacuated space.

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

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

    Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010 in Beijing, China. ... Mao Tsinghua University and Chair of the China Association for Hydrogen Energy 8:35 ...

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

    SciTech Connect (OSTI)

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

    1989-12-01

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

  15. Creep of A508/533 Pressure Vessel Steel

    SciTech Connect (OSTI)

    Richard Wright

    2014-08-01

    ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are

  16. Stress analysis and evaluation of a rectangular pressure vessel

    SciTech Connect (OSTI)

    Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel.

  17. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOE Patents [OSTI]

    Puechl, K.H.

    1963-09-24

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

  18. A DISLOCATION-BASED CLEAVAGE INITIATION MODEL FOR PRESSURE VESSEL

    SciTech Connect (OSTI)

    Cochran, Kristine B; Erickson, Marjorie A; Williams, Paul T; Klasky, Hilda B; Bass, Bennett Richard

    2012-01-01

    Efforts are under way to develop a theoretical, multi-scale model for the prediction of fracture toughness of ferritic steels in the ductile-to-brittle transition temperature (DBTT) region that accounts for temperature, irradiation, strain rate, and material condition (chemistry and heat treatment) effects. This new model is intended to address difficulties associated with existing empirically-derived models of the DBTT region that cannot be extrapolated to conditions for which data are unavailable. Dislocation distribution equations, derived from the theories of Yokobori et al., are incorporated to account for the local stress state prior to and following initiation of a microcrack from a second-phase particle. The new model is the basis for the DISlocation-based FRACture (DISFRAC) computer code being developed at the Oak Ridge National Laboratory (ORNL). The purpose of this code is to permit fracture safety assessments of ferritic structures with only tensile properties required as input. The primary motivation for the code is to assist in the prediction of radiation effects on nuclear reactor pressure vessels, in parallel with the EURATOM PERFORM 60 project.

  19. An investigation of RVACS (reactor vessel auxiliary cooling system) design improvements

    SciTech Connect (OSTI)

    Tzanos, C.P.; Tessier, J.H.; Pedersen, D.R. )

    1989-11-01

    One of the main safety features of the current liquid-metal reactor (LMR) designs is the utilization of decay heat removal systems that remove heat by natural convection. In the reactor vessel auxiliary cooling system (RVACS), decay heat is removed by naturally circulating air in the gap between the guard vessel and a baffle wall surrounding the guard vessel. The objective of this work was to determine the impact of a number of design parameters on the performance of the RVACS of a pool LMR. These parameters were (a) the stack height, (b) the size of the airflow gap, (c) the system pressure loss, (d) fins on the guard vessel or the baffle wall, and (e) roughness (in the form of repeated ribs) on the airflow channel walls. Reactor designs ranging from 400 to 3,500 MW(thermal) were considered. From the RVACS design parameters considered in this analysis, an optimized ribbed configuration gave the best improvement in RVACS performance. For a 3,500-MW(thermal) LMR, the peak sodium and cladding temperatures were reduced by 52 K.

  20. Report of the terawatt laser pressure vessel committee

    SciTech Connect (OSTI)

    Woodle, M.H.; Beauman, R.; Czajkowski, C.; Dickinson, T.; Lynch, D.; Pogorelsky, I.; Skjaritka, J.

    2000-09-25

    In 1995 the ATF project sent out an RFP for a CO2 Laser System having a TeraWatt output. Eight foreign and US firms responded. The Proposal Evaluation Panel on the second round selected Optoel, a Russian firm based in St. Petersburg, on the basis of the technical criteria and cost. Prior to the award, BNL representatives including the principal scientist, cognizant engineer and a QA representative visited the Optoel facilities to assess the company's capability to do the job. The contract required Optoel to provide a x-ray preionized high pressure amplifier that included: a high pressure cell, x-ray tube, internal optics and a HV pulse forming network for the main discharge and preionizer. The high-pressure cell consists of a stainless steel pressure vessel with various ports and windows that is filled with a gas mixture operating at 10 atmospheres. In accordance with BNL Standard ESH 1.4.1 ''Pressurized Systems For Experimental Use'', the pressure vessel design criteria is required to comply with the ASME Boiler and Pressure Vessel Code In 1996 a Preliminary Design Review was held at BNL. The vendor was requested to furnish drawings so that we could confirm that the design met the above criteria. The vendor furnished drawings did not have all dimensions necessary to completely analyze the cell. Never the less, we performed an analysis on as much of the vessel as we could with the available information. The calculations concluded that there were twelve areas of concern that had to be addressed to assure that the pressure vessel complied with the requirements of the ASME code. This information was forwarded to the vendor with the understanding that they would resolve these concerns as they continued with the vessel design and fabrication. The assembled amplifier pressure vessel was later hydro tested to 220 psi (15 Atm) as well as pneumatically to 181 psi (12.5 Atm) at the fabricator's Russian facility and was witnessed by a BNL engineer. The unit was shipped to the

  1. Cryogenic Pressure Vessels: Progress and Plans | Department of Energy

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

    Pressure Vessels: Progress and Plans Cryogenic Pressure Vessels: Progress and Plans Presented at the R&D Strategies for Compressed, Cryo-Compressed and Cryo-Sorbent Hydrogen Storage Technologies Workshops on February 14 and 15, 2011. compressed_hydrogen2011_9_aceves.pdf (1.88 MB) More Documents & Publications OEM Perspective on Cryogenic H2 Storage Cryo-Compressed Hydrogen Storage: Performance and Cost Review Proceedings of the 1998 U.S. DOE Hydrogen Program Review: April 28-30, 1998

  2. Partial Defect Testing of Pressurized Water Reactor Spent Fuel...

    Office of Scientific and Technical Information (OSTI)

    Partial Defect Testing of Pressurized Water Reactor Spent Fuel Assemblies Citation Details In-Document Search Title: Partial Defect Testing of Pressurized Water Reactor Spent Fuel ...

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

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07

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

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

    DOE Patents [OSTI]

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

    1993-01-01

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

  5. Threaded insert for compact cryogenic-capable pressure vessels

    DOE Patents [OSTI]

    Espinosa-Loza, Francisco; Ross, Timothy O.; Switzer, Vernon A.; Aceves, Salvador M.; Killingsworth, Nicholas J.; Ledesma-Orozco, Elias

    2015-06-16

    An insert for a cryogenic capable pressure vessel for storage of hydrogen or other cryogenic gases at high pressure. The insert provides the interface between a tank and internal and external components of the tank system. The insert can be used with tanks with any or all combinations of cryogenic, high pressure, and highly diffusive fluids. The insert can be threaded into the neck of a tank with an inner liner. The threads withstand the majority of the stress when the fluid inside the tank that is under pressure.

  6. CASL - PWR Reactor Vessel Multi-Physics CFD Model

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

    PWR Reactor Vessel Multi-Physics CFD Model Jin Yan*1, Yiban Xu1, Andrew Petrarca1, Zeses Karoutas1, Emre Tatli1, Emilio Baglietto2, Jess Gehin3 1Westinghouse Electric Company LLC 2Massachusetts Institute of Technology 3Oak Ridge National Lab *Correspondence to: yan3j@westinghouse.com A complete 3D SolidWorks CAD model of Watts Bar Unit 1 was constructed based on drawings. A single fuel assembly CAD model including all geometrical details was created based on the Westinghouse V5H 17x17 fuel

  7. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20

    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.

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

    SciTech Connect (OSTI)

    Elter, C.; Becker, G.

    1982-11-01

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

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

    SciTech Connect (OSTI)

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

    1997-04-01

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

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

    DOE Patents [OSTI]

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

    1997-01-01

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

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

    DOE Patents [OSTI]

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

    1997-12-16

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

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

    SciTech Connect (OSTI)

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

    1994-06-01

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

  13. Lightweight pressure vessels and unitized regenerative fuel cells

    SciTech Connect (OSTI)

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

    1996-12-31

    High specific energy (>400 Wh/kg) energy storage systems have been designed using lightweight pressure vessels in conjunction with unitized regenerative fuel cells (URFCs). URFCs produce power and electrolytically regenerate their reactants using a single stack of reversible cells. Although a rechargeable energy storage system with such high specific energy has not yet been fabricated, we have made progress towards this goal. A primary fuel cell (FC) test rig with a single cell (0.05 ft{sup 2} active area) has been modified and operated reversibly as a URFC. This URFC uses bifunctional electrodes (oxidation and reduction electrodes reverse roles when switching from charge to discharge, as with a rechargeable battery) and cathode feed electrolysis (water is fed from the oxygen side of the cell). Lightweight pressure vessels with state-of-the-art performance factors (burst pressure * internal volume/tank weight = Pb V/W) have been designed and fabricated. These vessels provide a lightweight means of storing reactant gases required for fuel cells (FCs) or URFCs. The vessels use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide the permeation barrier for gas storage. The bladders are fabricated using materials that are compatible with humidified gases which may be created by the electrolysis of water and are compatible with elevated temperatures that occur during fast fills.

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

    DOE Patents [OSTI]

    Lau, Louis K. S.

    1990-01-01

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

  15. Application of Negligible Creep Criteria to Candidate Materials for HTGR Pressure Vessels

    SciTech Connect (OSTI)

    Jetter, Robert I; Sham, Sam; Swindeman, Robert W

    2011-01-01

    Two of the proposed High Temperature Gas Reactors (HTGRs) under consideration for a demonstration plant have the design object of avoiding creep effects in the reactor pressure vessel (RPV) during normal operation. This work addresses the criteria for negligible creep in Subsection NH, Division 1 of the ASME B&PV (Boiler and Pressure Vessel) Code, Section III, other international design codes and some currently suggested criteria modifications and their impact on permissible operating temperatures for various reactor pressure vessel materials. The goal of negligible creep could have different interpretations depending upon what failure modes are considered and associated criteria for avoiding the effects of creep. It is shown that for the materials of this study, consideration of localized damage due to cycling of peak stresses results in a lower temperature for negligible creep than consideration of the temperature at which the allowable stress is governed by creep properties. In assessing the effect of localized cyclic stresses it is also shown that consideration of cyclic softening is an important effect that results in a higher estimated temperature for the onset of significant creep effects than would be the case if the material were cyclically hardening. There are other considerations for the selection of vessel material besides avoiding creep effects. Of interest for this review are (1) the material s allowable stress level and impact on wall thickness (the goal being to minimize required wall thickness) and (2) ASME Code approval (inclusion as a permitted material in the relevant Section and Subsection of interest) to expedite regulatory review and approval. The application of negligible creep criteria to two of the candidate materials, SA533 and Mod 9Cr-1Mo (also referred to as Grade 91), and to a potential alternate, normalized and tempered 2 Cr-1Mo, is illustrated and the relative advantages and disadvantages of the materials are discussed.

  16. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  17. File:06HIGBoilerPressureVesselPermit.pdf | Open Energy Information

    Open Energy Info (EERE)

    6HIGBoilerPressureVesselPermit.pdf Jump to: navigation, search File File history File usage Metadata File:06HIGBoilerPressureVesselPermit.pdf Size of this preview: 463 599...

  18. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect (OSTI)

    Hensel, S.

    2012-03-27

    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.

  19. Jam proof closure assembly for lidded pressure vessels

    DOE Patents [OSTI]

    Cioletti, Olisse C.

    1992-01-01

    An expendable closure assembly is provided for use (in multiple units) with a lockable pressure vessel cover along its rim, such as of an autoclave. This assembly is suited to variable compressive contact and locking with the vessel lid sealing gasket. The closure assembly consists of a thick walled sleeve insert for retention in the under bores fabricated in the cover periphery and the sleeve is provided with internal threading only. A snap serves as a retainer on the underside of the sleeve, locking it into an under bore retention channel. Finally, a standard elongate externally threaded bolt is sized for mating cooperation with the so positioned sleeve, whereby the location of the bolt shaft in the cover bore hole determines its compressive contact on the underlying gasket.

  20. Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

    SciTech Connect (OSTI)

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17

    Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing

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

    SciTech Connect (OSTI)

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

    1995-03-01

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

  2. A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

    SciTech Connect (OSTI)

    Chu, T.Y.; Bentz, J.; Simpson, R.; Witt, R.

    1997-02-01

    The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF program investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.

  3. A Survey of Pressure Vessel Code Compliance for Superconducting RF Cryomodules

    SciTech Connect (OSTI)

    Peterson, Thomas; Klebaner, Arkadiy; Nicol, Tom; Theilacker, Jay; Hayano, Hitoshi; Kako, Eiji; Nakai, Hirotaka; Yamamoto, Akira; Jensch, Kay; Matheisen, Axel; Mammosser, John; /Jefferson Lab

    2011-06-07

    Superconducting radio frequency (SRF) cavities made from niobium and cooled with liquid helium are becoming key components of many particle accelerators. The helium vessels surrounding the RF cavities, portions of the niobium cavities themselves, and also possibly the vacuum vessels containing these assemblies, generally fall under the scope of local and national pressure vessel codes. In the U.S., Department of Energy rules require national laboratories to follow national consensus pressure vessel standards or to show ''a level of safety greater than or equal to'' that of the applicable standard. Thus, while used for its superconducting properties, niobium ends up being treated as a low-temperature pressure vessel material. Niobium material is not a code listed material and therefore requires the designer to understand the mechanical properties for material used in each pressure vessel fabrication; compliance with pressure vessel codes therefore becomes a problem. This report summarizes the approaches that various institutions have taken in order to bring superconducting RF cryomodules into compliance with pressure vessel codes. In Japan, Germany, and the U.S., institutions building superconducting RF cavities integrated in helium vessels or procuring them from vendors have had to deal with pressure vessel requirements being applied to SRF vessels, including the niobium and niobium-titanium components of the vessels. While niobium is not an approved pressure vessel material, data from tests of material samples provide information to set allowable stresses. By means of procedures which include adherence to code welding procedures, maintaining material and fabrication records, and detailed analyses of peak stresses in the vessels, or treatment of the vacuum vessel as the pressure boundary, research laboratories around the world have found methods to demonstrate and document a level of safety equivalent to the applicable pressure vessel codes.

  4. Improved mechanical properties of A 508 class 3 steel for nuclear pressure vessel through steelmaking

    SciTech Connect (OSTI)

    Kim, J.T.; Kwon, H.K.; Kim, K.C.; Kim, J.M.

    1997-12-31

    The present work is concerned with the steelmaking practices which improve the mechanical properties of the A 508 class 3 steel for reactor pressure vessel. Three kinds of steelmaking practices were applied to manufacture the forged heavy wall shell for reactor pressure vessel, that is, the vacuum carbon deoxidation (VCD), modified VCD containing aluminum and silicon-killing. The segregation of the chemical elements through the thickness was quite small so that the variations of the tensile properties at room temperature were small and the anisotropy of the impact properties was hardly observed regardless of the steelmaking practices. The Charpy V-notch impact properties and the reference nil-ductile transition temperature by drop weight test were significantly improved by the modified VCD and silicon-killing as compared with those of the steel by VCD. Moreover, the plane strain fracture toughness values of the materials by modified VCD and silicon-killing practices was much higher than those of the steel by VCD. These were resulted from the fining of austenite grain size. It was observed that the grain size was below 20 {micro}m (ASTM No. 8.5) when using the modified VCD and silicon-killing, compared to 50 {micro}m (ASTM No. 7.0) when using VCD.

  5. Structural design, analysis, and code evaluation of an odd-shaped pressure vessel

    SciTech Connect (OSTI)

    Rezvani, M.A.; Ziada, H.H.

    1992-12-01

    This paper is the result of an effort to design, analyze and evaluate a rectangular pressure vessel. Normally pressure vessels are designed in circular or spherical shapes to prevent stress concentrations. In this case, because of operational limitations, the choice of vessels was limited to a rectangular pressure box with a removable cover plate. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code is used as a guideline for pressure containments whose width or depth exceeds 15.24 cm (6.0 in.) and where pressures will exceed 103.4 KPa (15.0 lbf/in[sup 2]). This evaluation used Section VIII of this Code, hereafter referred to as the Code. The dimensions and working pressure of the subject vessel fall within the pressure vessel category of the Code. The Code design guidelines and rules do not directly apply to this vessel. Therefore, finite-element methodology was used to analyze the pressure vessel, and the Code then was used in qualifying the vessel to be stamped to the Code. Section VIII, Division 1 of the Code was used for evaluation. This action was justified by selecting a material for which fatigue damage would not be a concern. The stress analysis results were then chocked against the Code, and the thicknesses adjusted to satisfy Code requirements. Although not directly applicable, the Code design formulas for rectangular vessels were also considered and presented in this study.

  6. Structural design, analysis, and code evaluation of an odd-shaped pressure vessel

    SciTech Connect (OSTI)

    Rezvani, M.A.; Ziada, H.H.

    1992-12-01

    This paper is the result of an effort to design, analyze and evaluate a rectangular pressure vessel. Normally pressure vessels are designed in circular or spherical shapes to prevent stress concentrations. In this case, because of operational limitations, the choice of vessels was limited to a rectangular pressure box with a removable cover plate. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code is used as a guideline for pressure containments whose width or depth exceeds 15.24 cm (6.0 in.) and where pressures will exceed 103.4 KPa (15.0 lbf/in{sup 2}). This evaluation used Section VIII of this Code, hereafter referred to as the Code. The dimensions and working pressure of the subject vessel fall within the pressure vessel category of the Code. The Code design guidelines and rules do not directly apply to this vessel. Therefore, finite-element methodology was used to analyze the pressure vessel, and the Code then was used in qualifying the vessel to be stamped to the Code. Section VIII, Division 1 of the Code was used for evaluation. This action was justified by selecting a material for which fatigue damage would not be a concern. The stress analysis results were then chocked against the Code, and the thicknesses adjusted to satisfy Code requirements. Although not directly applicable, the Code design formulas for rectangular vessels were also considered and presented in this study.

  7. The criteria of fracture in the case of the leak of pressure vessels

    SciTech Connect (OSTI)

    Habil; Ziliukas, A.

    1997-04-01

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  8. D-Zero Central Calorimeter Pressure Vessel and Vacuum Vessel Safety Notes

    SciTech Connect (OSTI)

    Rucinski, R.; Luther, R.; /Fermilab

    1990-10-25

    The relief valve and relief piping capacity was calculated to be 908 sefm air. This exceeds all relieving conditions. The vessel also has a rupture disc with a 2640 scfm air stamped capacity. In order to significantly decrease the amount of time required to fill the cryostats, it is desired to raise the setpoint of the 'operating' relief valve on the argon storage dewar to 20 psig from its existing 16 psig setting. This additional pressure increases the flow to the cryostats and will overwhelm the relief capacity if the temperature of the modules within these vessels is warm enough. Using some conservative assumptions and simple calculations within this note, the maximum average temperature that the modules within each cryostat can be at prior to filling from the storage dewar with liquid argon is at least 290 K. The average temperature of the module mass for any of the three cryostats can be as high as 290 K prior to filling that particular cryostat. This should not be confused with the average temperature of a single type or location which is useful in protecting the modules-not necessarily the vessel itself. A few modules of each type and at different elevations should be used in an average which would account for the different weights of each module. Note that at 290 K, the actual flow of argon through the relief valve and the rupture disk was under the maximum theoretical flows for each relief device. This means that the bulk temperature could actually have been raised to flow argon through the reliefs at their maximum capacity. Therefore, the temperature of 290 K is a conservative value for the calculated flow rate of 12.3 gpm. Safeguards in addition to and used in conjunction with operating procedures shall be implemented in such a way so that the above temperature limitation is not exceeded and such that it is exclusive of the programmable logic controller (PLC). One suggestion is using a toggle switch for each cryostat mounted in the PLC I/O box which

  9. CONVECTION REACTOR

    DOE Patents [OSTI]

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

    1960-03-22

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

  10. Cleavage Fracture Modeling of Pressure Vessels under Transient Thermo-Mechanical Loading

    SciTech Connect (OSTI)

    Qian, Xudong; Dodds, Robert; Yin, Shengjun; Bass, Bennett Richard

    2008-02-01

    The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models become appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from the

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

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

    2010-06-07

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

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1995-01-01

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

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

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

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

  15. Fuel assembly transfer basket for pool type nuclear reactor vessels

    DOE Patents [OSTI]

    Fanning, Alan W.; Ramsour, Nicholas L.

    1991-01-01

    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.

  16. Stress-intensity-factor influence coefficients for semielliptical inner-surface flaws in clad pressure vessels

    SciTech Connect (OSTI)

    Keeney, J.A.; Bryson, J.W.

    1995-12-31

    A problem of particular interest in pressure vessel technology is the calculation of accurate stress-intensity factors for semielliptical surface cracks in cylinders. Computing costs for direct solution techniques can be prohibitive when applied to three-dimensional (3-D) geometries with time-varying boundary conditions such as those associated with pressurized thermal shock. An alternative superposition technique requires the calculation of a set of influence coefficients for a given 3-D crack model that can be superimposed to obtain mode-I stress-intensity factors. This paper presents stress-intensity-factor influence coefficients (SIFICs) for axially and circumferentially oriented finite-length semielliptical inner-surface flaws with aspect ratios (total crack length (2c) to crack depth (a)) of 2, 6, and 10 for clad cylinders having an internal radius to wall thickness (t) ratio of 10. SIFICs are computed for flaw depths in the range of 0.01 {le} a/t {le} 0.5 and two cladding thicknesses. The incorporate of this SIFIC data base in fracture mechanics codes will facilitate the generation of fracture mechanics solutions for a wide range of flaw geometries as may be required in structural integrity assessments of pressurized-water and boiling-water reactors.

  17. Photoacoustic sample vessel and method of elevated pressure operation

    DOE Patents [OSTI]

    Autrey, Tom; Yonker, Clement R.

    2004-05-04

    An improved photoacoustic vessel and method of photoacoustic analysis. The photoacoustic sample vessel comprises an acoustic detector, an acoustic couplant, and an acoustic coupler having a chamber for holding the acoustic couplant and a sample. The acoustic couplant is selected from the group consisting of liquid, solid, and combinations thereof. Passing electromagnetic energy through the sample generates an acoustic signal within the sample, whereby the acoustic signal propagates through the sample to and through the acoustic couplant to the acoustic detector.

  18. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  19. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20

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

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

    SciTech Connect (OSTI)

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

    2010-04-13

    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.

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

    SciTech Connect (OSTI)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. ); Nichols, R.T. ); Sweet, D.W. )

    1991-08-01

    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.

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

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

    These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpvlynch.pdf (4.21 MB) More Documents & ...

  3. Specifications for the development of BUGLE-93: An ENDF/B-VI multigroup cross section library for LWR shielding and pressure vessel dosimetry

    SciTech Connect (OSTI)

    White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.

    1992-11-01

    This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.

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

    SciTech Connect (OSTI)

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

    2005-01-01

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

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

    SciTech Connect (OSTI)

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

    2010-11-10

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

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

    SciTech Connect (OSTI)

    Langton, C.; Stefanko, D.

    2011-01-05

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

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

    DOE Patents [OSTI]

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

    1988-01-01

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

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

    SciTech Connect (OSTI)

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

    2014-07-10

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

  9. OVERVIEW OF PRESSURE VESSEL DESIGN CRITERIA FOR INTERNAL DETONATION (BLAST) LOADING

    SciTech Connect (OSTI)

    T. A. DUFFEY; E. A. RODRIGUEZ

    2001-05-01

    Spherical and cylindrical pressure vessels are often used to completely contain the effects of high explosions. These vessels generally fall into two categories. The first includes vessels designed for multiple use ([1]-[6]). Applications of such multiple-use vessels include testing of explosive components and bomb disposal. Because of the multiple-use requirement, response of the vessel is restricted to the elastic range. The second category consists of vessels designed for one-time use only ([7]-[9]). Vessels in this category are typically used to contain accidental explosions and are designed to efficiently utilize the significant plastic energy absorption capacity of ductile materials. Because these vessels may undergo large permanent plastic deformations, they may not be reusable. Ideally one would design a Containment Vessel according to some National or International Consensus Standard, such as the ASME Boiler and Pressure Vessel Code. Unfortunately, however, a number of issues preclude direct use of the ASME Code in its present form to the design of Containment Vessels. These issues are described in Section 2, along with a request for guidance from the PVRC as to a suitable path forward for developing appropriate ASME B&PV design guidance for Containment Vessels. Next, a discussion of the nature of impulsive loading as a result of an internal detonation of the high explosive within a Containment Vessel is described in Section 3. Ductile failure criteria utilized for LANL Containment Vessels are described in Section 4. Finally, brittle fracture criteria currently utilized by LANL are presented in Section 5. This memo is concluded with a brief summary of results and an appeal to PVRC to recommend and develop an appropriate path forward (Section 6). This path forward could be of a short-term specialized nature (e.g., Code Case) for specific guidance regarding design of the LANL Containment Vessels; a long-term development of a general design approach

  10. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, F.E.

    1992-12-08

    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.

  11. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, Franklin E.

    1992-01-01

    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.

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

    SciTech Connect (OSTI)

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

    1981-01-01

    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.

  13. Propellant actuated nuclear reactor steam depressurization valve

    DOE Patents [OSTI]

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

    1992-01-01

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

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

    SciTech Connect (OSTI)

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

    2013-11-26

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

  15. Development of design criteria for a high pressure vessel construction code

    SciTech Connect (OSTI)

    Mraz, G.J.

    1987-05-01

    Out of concern for public safety, most legal jurisdictions now require unfired pressure vessel construction to comply with the ASME Boiler and Pressure Vessel Code. Because the present two divisions of Section VIII of that Code are not well suited for high pressure design, a new division is needed. The currently anticipated main design criteria of the proposed division are full plastic flow or full overstrain pressure, stress intensity in the bore, fatigue, and fracture mechanics. The rules are expected to allow better utilization of high strength steels already included in the present Section VIII. At the same time materials of even higher strength are introduced. The benefits of compressive prestress are recognized. Construction methods allowing it's achievement, such as autofrettage, shrink fitting and wire winding are included. Reasons for selection of the criteria are given.

  16. Control of reactor coolant flow path during reactor decay heat removal

    DOE Patents [OSTI]

    Hunsbedt, Anstein N.

    1988-01-01

    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.

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

    SciTech Connect (OSTI)

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

    1994-06-01

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

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

    SciTech Connect (OSTI)

    Rana, M.D.; Zawierucha, R.

    1995-12-01

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

  19. Pressure surge attenuator

    DOE Patents [OSTI]

    Christie, Alan M.; Snyder, Kurt I.

    1985-01-01

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

  20. A Multiscale Modeling Approach to Analyze Filament-Wound Composite Pressure Vessels

    SciTech Connect (OSTI)

    Nguyen, Ba Nghiep; Simmons, Kevin L.

    2013-07-22

    A multiscale modeling approach to analyze filament-wound composite pressure vessels is developed in this article. The approach, which extends the Nguyen et al. model [J. Comp. Mater. 43 (2009) 217] developed for discontinuous fiber composites to continuous fiber ones, spans three modeling scales. The microscale considers the unidirectional elastic fibers embedded in an elastic-plastic matrix obeying the Ramberg-Osgood relation and J2 deformation theory of plasticity. The mesoscale behavior representing the composite lamina is obtained through an incremental Mori-Tanaka type model and the Eshelby equivalent inclusion method [Proc. Roy. Soc. Lond. A241 (1957) 376]. The implementation of the micro-meso constitutive relations in the ABAQUS finite element package (via user subroutines) allows the analysis of a filament-wound composite pressure vessel (macroscale) to be performed. Failure of the composite lamina is predicted by a criterion that accounts for the strengths of the fibers and of the matrix as well as of their interface. The developed approach is demonstrated in the analysis of a filament-wound pressure vessel to study the effect of the lamina thickness on the burst pressure. The predictions are favorably compared to the numerical and experimental results by Lifshitz and Dayan [Comp. Struct. 32 (1995) 313].

  1. Upper internals arrangement for a pressurized water reactor

    DOE Patents [OSTI]

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

    2013-07-09

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

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

    DOE Patents [OSTI]

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

    2013-01-15

    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.

  3. Pipeline and Pressure Vessel R&D under the Hydrogen Regional Infrastructure Program In Pennsylvania

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

    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, SC Acknowledgments * This material is based upon work supported by the Department of Energy under Award Number DE-FC36-04GO14229 * Partners - Savannah River National Laboratory (SRNL) - HyPerComp Engineering Inc. (HEI) - American Society Of Mechanical Engineers (ASME) - Pipeline Working Group (PWG) Program

  4. Autogenic pressure reactors provide simple, rapid means of producing

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

    battery materials - Energy Innovation Portal Energy Storage Energy Storage Advanced Materials Advanced Materials Find More Like This Return to Search Autogenic pressure reactors provide simple, rapid means of producing battery materials Argonne National Laboratory Contact ANL About This Technology Spherical carbon particles prepared in an autogenic reaction Spherical carbon particles prepared in an autogenic reaction Technology Marketing Summary Rechargeable lithium-ion batteries have become

  5. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOE Patents [OSTI]

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

    1960-08-16

    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.

  6. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

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

    DOE Patents [OSTI]

    Milligan, John D.

    1983-01-01

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

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

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-12-29

    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

  9. Fundamental underwater cutting method experiment as a dismantling tool for a commercial atomic reactor vessel

    SciTech Connect (OSTI)

    Hamasaki, M.; Murao, Y.; Tateiwa, F.

    1982-10-01

    A new underwater cutting technique applying underwater dismantling to commercial atomic reactor vessels has been developed. This technique involves gas cutting the mild steel underwater after removing the stainless steel cladding by arc gouging. The arc gouging is achieved by blowing out metal--which is melted by an arc between a mild steel electrode wire and the stainless steel--by jetting water from a rear water nozzle. The fuel gas employed for preheating for the gas cutting was a mixed gas of propane and 30% methylacetylene. The test piece used was made of 300-mm-thick mild steel with 8-mm-thick stainless steel cladding. The fundamental cutting experiment was carried out successfully under a cutting speed condition of 15 cm/min at a water depth of 20 cm. This apparatus is easy to handle, compact, and cheap.

  10. Measurement of Fatigue Crack Growth Relationships in Hydrogen Gas for Pressure Swing Adsorber Vessel Steels

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Somerday, Brian P.; Barney, Monica

    2014-12-04

    We measured the hydrogen-assisted fatigue crack growth rates (da/dN) for SA516 Grade 70 steel as a function of stress-intensity factor range (ΔK) and load-cycle frequency to provide life-prediction data relevant to pressure swing adsorber (PSA) vessels. For ΔK values up to 18.5 MPa m1/2, the baseline da/dN versus ΔK relationship measured at 1Hz in 2.8 MPa hydrogen gas represents an upper bound with respect to crack growth rates measured at lower frequency. However, at higher ΔK values, we found that the baseline da/dN data had to be corrected to account for modestly higher crack growth rates at the lower frequenciesmore » relevant to PSA vessel operation.« less

  11. Measurement of Fatigue Crack Growth Relationships in Hydrogen Gas for Pressure Swing Adsorber Vessel Steels

    SciTech Connect (OSTI)

    Somerday, Brian P.; Barney, Monica

    2014-12-04

    We measured the hydrogen-assisted fatigue crack growth rates (da/dN) for SA516 Grade 70 steel as a function of stress-intensity factor range (ΔK) and load-cycle frequency to provide life-prediction data relevant to pressure swing adsorber (PSA) vessels. For ΔK values up to 18.5 MPa m1/2, the baseline da/dN versus ΔK relationship measured at 1Hz in 2.8 MPa hydrogen gas represents an upper bound with respect to crack growth rates measured at lower frequency. However, at higher ΔK values, we found that the baseline da/dN data had to be corrected to account for modestly higher crack growth rates at the lower frequencies relevant to PSA vessel operation.

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

    SciTech Connect (OSTI)

    Cannell, Gary L.; Huth, Ralph J.; Hallum, Randall T.

    2013-08-26

    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.

  13. D-Zero Central Calorimeter Technical Appendix to Cryogenic Pressure Vessels

    SciTech Connect (OSTI)

    Mulholland, G.T.; Rucinski, R.A.; /Fermilab

    1990-11-19

    DO (D Zero) is a large Liquid Argon (LAr) HEP Calorimeter designed to function in the laboratories P-Pbar collider at the DO section of the Tevatron accelerator. It contains 5,000 gls. of LAr in the CC cryostat, and 3,000 gls. in each of two, a north and south, EC cryostats. These low pressure vessels are filled with detector modules built of stainless steel, copper and depleted uranium. The LAr functions as the ionization medium, and the spatial and temporal of the collection of the charge of the electrons produced signals the passsage of charged particles. The collection of these charges in 4 pi is related to the energy of the particles, and their measurement is called calorimetry. The contained LAr (T=90K) is isolated from the ambient temperatures in specially designed, vacuum and superinsulated, vessels (cryostats) provided with liquid nitrogen, heat of vaporization, cooling.

  14. Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

    SciTech Connect (OSTI)

    Naus, D.J

    1983-01-01

    Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development.

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

    SciTech Connect (OSTI)

    Rana, M.D. . Research and Development Dept.)

    1994-08-01

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

  16. Low pressure stagnation flow reactor with a flow barrier

    DOE Patents [OSTI]

    Vosen, Steven R.

    2001-01-01

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

  17. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    SciTech Connect (OSTI)

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

  18. Welding and brazing qualifications (supplement to ASME Boiler and Pressure Vessel Code, Section IX)

    SciTech Connect (OSTI)

    Not Available

    1981-11-01

    This standard supplements the requirements of the 1980 edition of the ASME Boiler and Pressure Vessel Code (the Code), Section IX. When this standard is invoked or referenced, the applicable subsections of Section IX of the Code are also invoked or referenced. The paragraph numbers in this standard apply only to the 1980 edition of Section IX and its addenda. The user of this standard is responsible for obtaining and applying the edition and revision of this standard that supplement the edition and addenda of Section IX that are in legal effect at the time of use.

  19. Welding and brazing qualifications (supplement to ASME boiler and pressure vessel code, Section IX)

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    This standard supplements the requirements of the 1977 edition of the ASME Boiler and Pressure Vessel Code (the Code), Section IX. When this standard is invoked or referenced, the applicable subsections of Section IX of the Code are also invoked or referenced. The paragraph numbers apply only to the 1977 edition of Section IX and its addenda. The user of this standard is responsible for obtaining and applying the edition and revision of this standard that supplement the edition and Addenda of Section IX that are in legal effect at the time of use.

  20. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    SciTech Connect (OSTI)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

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

    SciTech Connect (OSTI)

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

    1997-04-01

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

  2. FOOD IRRADIATION REACTOR

    DOE Patents [OSTI]

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

    1961-05-01

    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.

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

    SciTech Connect (OSTI)

    Langton, C.; Stefanko, D.

    2011-03-10

    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].

  4. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

    Office of Environmental Management (EM)

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

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

    SciTech Connect (OSTI)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  6. Effect of compression on individual pressure vessel nickel/hydrogen components

    SciTech Connect (OSTI)

    Manzo, M.A.; Perez-Davis, M.E.

    1988-08-01

    Compression tests were performed on representative Individual Pressure Vessel (IPV) Nickel/Hydrogen cell components in an effort to better understand the effects of force on component compression and the interactions of components under compression. It appears that the separator is the most easily compressed of all of the stack components. It will typically partially compress before any of the other components begin to compress. The compression characteristics of the cell components in assembly differed considerably from what would be predicted based on individual compression characteristics. Component interactions played a significant role in the stack response to compression. The results of the compression tests were factored into the design and selection of Belleville washers added to the cell stack to accommodate nickel electrode expansion while keeping the pressure on the stack within a reasonable range of the original preset.

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

    SciTech Connect (OSTI)

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

    2009-10-15

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

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

    SciTech Connect (OSTI)

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

    2015-05-15

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

  9. Specialist meeting on leak before break in reactor piping and vessels

    SciTech Connect (OSTI)

    Bartholome, G.; Bazant, E.; Wellein, R.

    1997-04-01

    A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

  10. Effect of fins and repeated-rib roughness on the performance characteristics of a reactor vessel air cooling system for LMFBR shutdown heat removal

    SciTech Connect (OSTI)

    Cheung, F.B.; Chawla, T.C.; Pedersen, D.R.; Tessier, J.H.; Webb, R.L.

    1986-01-01

    The use of a totally passive cooling system for shutdown heat removal that rejects heat from the reactor vessel by radiation to the guard vessel and from the guard vessel to a circulating air stream driven by natural convection is a key feature of the US Department of Energy's liquid-metal reactor advanced design study concepts. General Electric refers to the system as the Reactor Vessel Auxiliary Cooling System (RVACS) and Rockwell International as the Reactor Auxiliary Cooling System (RACS). The circulating air stream is contained in the annular passage formed with guard vessel wall and the duct wall surrounding the guard vessel. Specifically, the RVACS/RACS is designed to assure adequate cooling of the reactor vessel under abnormal operational conditions associated with loss of heat removal through the normal heat transport path via the steam generator system or the DRACS, if available. To enhance the heat transfer, longitudinal radial fins or repeated ribs can be attached to the duct wall and/or the guard vessel. The purpose of the present paper is to summarize the status of the analytical work on the development of an optimum design configuration for the RVACS/RACS.

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

    SciTech Connect (OSTI)

    Luk, K.H.

    1993-09-01

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

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

    SciTech Connect (OSTI)

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

    2012-09-01

    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

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

    SciTech Connect (OSTI)

    Wiersma, B.

    2010-05-24

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

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

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

    Department of Energy 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

  16. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

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

  17. Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

    SciTech Connect (OSTI)

    Shwageraus, Eugene; Zhao Xianfeng; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J. Stephen

    2004-07-15

    A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with {sup 235}U is necessary, and the {sup 235}U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO{sub 2}-UO{sub 2}) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of {sup 233}U from the {sup 232}Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the {sup 233}U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible

  18. Using SA508/533 for the HTGR Vessel Material

    SciTech Connect (OSTI)

    Larry Demick

    2012-06-01

    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.

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

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-10-29

    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

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

    SciTech Connect (OSTI)

    Gat, U.; Kasten, P.R.

    1980-11-01

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

  1. Overview of new rules and recent changes in ASME code, Section VIII, pressure vessels

    SciTech Connect (OSTI)

    Farr, J.R.

    1995-12-01

    In this presentation, some of the new rules and recent changes to the ASME Boiler and Pressure Vessel Code, Section 8, Divisions 1 and 2, are reviewed. On July 1, 1995, the 1995 Edition of the ASME Code was issued. This 1995 Edition incorporates those items which were added of changed in the 1992, 1993, and 1994 Addenda to the 1992 Edition of the Code. The 1995 Edition contains no new items which were not included in the previous edition and three addenda. With the possibility of an extended time before some of the new rules are able to appear in the addenda, the recent trend is to put the rules in Code Cases which are approved earlier. Consequently, it is necessary to review new Code Cases as well as Code changes. Updates continue for impact requirements for standard components as well as for materials other than UCS, carbon steel and low alloys. Extensive changes have been made for UHA, high-alloy, materials regarding impact requirements. Example problems have been revised to include these effects. Significant changes are reviewed.

  2. Cryogenic Pressure Vessels for H2 Vehicles Rapidly Refueled by LH2 pump to 700 bar

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

    Vessels for H 2 Vehicles Rapidly Refueled by LH 2 pump to 700 bar Salvador Aceves, Gene Berry, Guillaume Petitpas, Vernon Switzer Lawrence Livermore National Laboratory CAMX meeting October 29 th , 2015 LLNL-PRES-678629 * Cryogenic H 2 Onboard Storage * Temperature as a Degree of Freedom in H 2 storage * LLNL Cryocompressed Project History * 350 Bar Test Vehicle Park & Drive Results * Current Project * 700 bar prototype (cryogenic) vessels * Refueling with LH 2 Pump * Test Vessel Cycling

  3. Life assessment of a C-1/2Mo petroleum refinery pressure vessel operating in the creep regime

    SciTech Connect (OSTI)

    Brown, R.G.; Osage, D.A.; Buchheim, G.M.; Dobis, J.D.

    1995-12-31

    A comprehensive fitness-for-service assessment was conducted to evaluate a C-1/2Mo pressure vessel which has operated at temperatures in the creep range for almost 45 years. An initial damage assessment based on elastic stress analysis results indicated that this vessel was approaching its predicted failure life and thus there was little potential for increasing the operating temperature. Creep tests were conducted on samples removed from high stress regions of the vessel according to the MPC Omega Program protocol. The creep test results indicated that the material possesses creep strength superior to average new material and therefore has substantial remaining life. A nonlinear finite element analysis incorporating the MPC Project Omega creep law was performed to assess creep and fatigue damage. The results of this assessment indicated that future operation at increased temperatures was indeed feasible.

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

    DOE Patents [OSTI]

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

    1994-05-03

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

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

    DOE Patents [OSTI]

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

    1994-01-01

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

  6. Ceramic membrane reactor with two reactant gases at different pressures

    DOE Patents [OSTI]

    Balachandran, Uthamalingam; Mieville, Rodney L.

    2001-01-01

    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.

  7. Review of the state of criticality of the Three Mile Island Unit 2 core and reactor vessel

    SciTech Connect (OSTI)

    Stratton, W.R. )

    1987-04-15

    The events during the early hours of the Three Mile Island Unit 2 (TMI-2) accident on March 28, 1979 caused the fuel in the reactor core to crumble or disintegrate, and then subside into a rubble structure more compact that its normal configuration. The present height of the core is about seven feet, five feet less than its normal configuration of 12 feet. With the same boron content and some or all of the control rod and burnable poison rod material as the normal core configuration, the collapsed structure is calculated to be more reactive. However, the reactor is assuredly subcritical at present because of the extraordinarily high boron concentration maintained in the coolant water. Four additional and different physical models are discussed briefly in the report to illustrate the margin of subcriticality, to provide a better estimate of the neutron multiplication factor, and to provide some understanding of the criticality effects of the important parameters. Two different finite, cylindrical models of a collapsed core are also presented in this report. The conclusion of this review is that the reactor is now very far subcritical with a boron concentration of 4350 ppM or more, and no conceivable rearrangement of fuel can create a critical state. Careful administrative control to maintain the boron concentration of the reactor coolant close to 5000 ppM, and controls to rigorously exclude addition of unborated water to the primary system, provide additional assurance that subcriticality will be maintained. The immediate corollary is that the defueling of the reactor vessel can proceed as planned, with complete confidence that such operations will remain subcritical. 20 refs.

  8. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    SciTech Connect (OSTI)

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers & Constructors and Chicago Bridge & Iron (Raytheon/CB&I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB&I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB&I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB&I and documented accordingly.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1958-12-01

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

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

    SciTech Connect (OSTI)

    Bhuyan, G.S.; Sperling, E.J.; Shen, G.; Yin, H.; Rana, M.D.

    1996-12-01

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

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

    SciTech Connect (OSTI)

    Bhuyan, G.S.; Sperling, E.J.; Shen, G.; Yin, H.; Rana, M.D.

    1999-08-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    1997-04-01

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

  14. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  15. Design practices in Japan for the super high pressure vessels and comparison with the ASME Code Sect. VIII Div. 3 (under preparation)

    SciTech Connect (OSTI)

    Onozawa, Tsutomu; Tahara, Takayasu

    1995-12-01

    Recently, super high pressure facilities have been increasing in the industrial area so that to establish the regulatory standard to regulate the super high pressure vessels is a matter of great urgency world widely to keep the industrial safety. Under such a situation, the author shows respect to the ASME Code Committee for their efforts to publish the super high pressure vessel code. Mr. Leslie P. Antalffy, Fluor Daniel, Incorporated, Houston, Texas presented a paper during the 1993 and 1994 ASME PVP Conferences that ASME Code Committee has been preparing the rules of Division 3 of Section 8 of the Boiler and Pressure Vessel Code and explained its outline. In this paper, the authors shows the current super high pressure vessel design practices in Japan and explain the merit and problem area of these formulas comparing with the ASME formula and necessary conditions for the fatigue analysis.

  16. Detailed Analysis of a Late-Phase Core-Melt Progression for the Evaluation of In-vessel Corium Retention

    SciTech Connect (OSTI)

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

    2006-12-01

    Detailed analyses of a late-phase melt progression in the advanced power reactor (APR)1400 were completed to identify the melt and the thermal-hydraulic states of the in-vessel materials in the reactor vessel lower plenum at the time of reactor vessel failure to evaluate the candidate strategies for an in-vessel corium retention (IVR). Initiating events considered included high-pressure transients of a total loss of feed water (LOFW) and a station blackout (SBO) and low-pressure transients of a 0.0009-m2 small, 0.0093-m2 medium, and 0.0465-m2 large-break loss-of-coolant accident (LOCA) without safety injection. Best-estimate simulations for these low-probability events with conservative accident progression assumptions that lead to reactor vessel failure were performed by using the SCDAP/RELAP5/MOD3.3 computer code. The SCDAP/RELAP5/MOD3.3 results have shown that the pressurizer surge line failed before the reactor vessel failure, which results in a rapid decrease of the in-vessel pressure and a delay of the reactor vessel failure time of ~40 min in the high-pressure sequences of the total LOFW and the SBO transients. In all the sequences, ~80 to 90% of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. The maximum value of the volumetric heat source in the corium pool was estimated as 1.9 to 3.7 MW/m3. The corium temperature was ~2800 to 3400 K at the time of reactor vessel failure. The highest volumetric heat source sequence is predicted for the 0.0465-m2 large-break LOCA without safety injection in the APR1400, because this sequence leads to an early reactor vessel failure.

  17. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

    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.

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

    SciTech Connect (OSTI)

    Michael L. Swanson

    2005-08-30

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

  19. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2007-02-20

    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.

  20. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2008-02-26

    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.

  1. Ion transport membrane module and vessel system

    DOE Patents [OSTI]

    Stein, VanEric Edward; Carolan, Michael Francis; Chen, Christopher M.; Armstrong, Phillip Andrew; Wahle, Harold W.; Ohrn, Theodore R.; Kneidel, Kurt E.; Rackers, Keith Gerard; Blake, James Erik; Nataraj, Shankar; Van Doorn, Rene Hendrik Elias; Wilson, Merrill Anderson

    2012-02-14

    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.

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

    SciTech Connect (OSTI)

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

    2003-04-01

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

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

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

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

  4. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    SciTech Connect (OSTI)

    Daily, Charles R.

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  5. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J.

    1987-01-01

    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.

  6. Pressure suppression containment system for boiling water reactor

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

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

  7. Pressure suppression containment system for boiling water reactor

    DOE Patents [OSTI]

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

    1997-01-21

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

  8. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

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

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

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

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

    SciTech Connect (OSTI)

    Rawls, G

    2007-04-30

    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.

  11. Optimization of Pressurized Oxy4Combustion with Flameless Reactor

    SciTech Connect (OSTI)

    Malavasi, Massimo; Landegger, Gregory

    2014-06-30

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

  12. Residence time distribution studies in a multiphase reactor under high temperature and pressure conditions

    SciTech Connect (OSTI)

    Nalitham, R.V.; Davies, O.L.

    1987-06-01

    The residence time distribution of the slurry phase in a coal liquefaction reactor is determined experimentally under high temperature and pressure conditions using native radioactive tracers. The experimental data are fitted to several exit age distribution models, and a model is selected based on the best fit. The effect of process conditions such as recycle gas rate, coal feed rate, reactor temperature, and solvent-to-coal ratio on the degree of backmixing and mean residence time is studied. Gas holdup is estimated from the experimental mean residence time, the nominal residence time, and the total reactor holdup. The effect of gas superficial velocity on gas holdup is studied.

  13. Component failures at pressurized water reactors. Final report

    SciTech Connect (OSTI)

    Reisinger, M.F.

    1980-10-01

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

  14. Creep behavior of a nuclear pressure vessel under severe accident conditions

    SciTech Connect (OSTI)

    Beghini, M.; Bertini, L.; Vitale, E.

    1996-12-31

    The results of a study on the creep behavior of the vessel lower head under severe accident conditions are reported. An experimental program aimed at the evaluation of the creep properties of A533grB steel at high temperature (800--1,100 C) and under biaxial loading is summarized and the main results reported. A Finite Element simulation of the lower head under severe accident conditions allows to show the effect of the main parameters affecting the time to rupture.

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

    SciTech Connect (OSTI)

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

    2015-01-01

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

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

    SciTech Connect (OSTI)

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

    2015-01-01

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

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

    DOE Patents [OSTI]

    Efthymiou, George S. (Limassol, CY); Shuler, Michael L. (Ithaca, NY)

    1989-08-29

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

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

    DOE Patents [OSTI]

    Hill, Paul R.

    1994-01-01

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

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

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27

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

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

    SciTech Connect (OSTI)

    Yongfeng Zhang; Pritam Chakraborty; S. Bulent Biner

    2013-09-01

    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.

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  2. Effect of silicon on ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels

    SciTech Connect (OSTI)

    Qiu, J.A.; Wu, K.M.; Li, J.H.; Hodgson, P.D.; Hou, T.P.; Ding, Q.F.

    2013-09-15

    The effect of Si on the ultra-low temperature toughness of Nb–Ti microalloyed cryogenic pressure vessel steels was investigated by electron back-scattered diffraction and transmission electron microscope with energy dispersive spectroscopy. Equiaxed ferrite and bainite were obtained in the tempered steels with small Si additions. Nanosized Nb–Ti carbides (< 10 nm) were formed in the steel containing 0.05% Si, whereas much coarser carbides (> 30 nm) were found in the steel containing 0.47% Si. The ultra-low temperature toughness of the Nb–Ti microalloyed cryogenic pressure vessel steel was remarkably enhanced by the reduction in the Si content, which was attributed to the pre-existing iron carbide formation before the precipitation of nanosized Nb–Ti carbides during tempering. - Highlights: • Nanosized Nb-Ti carbides formed in the tempered steel with smaller Si addition. • Coarser Nb-Ti carbides formed in the tempered steel with more Si addition. • Pre-existing cememtites provide nucleation sites for Nb-Ti carbide precipitation. • Ultra-low temperature toughness was remarkably enhanced by Si content reduction.

  3. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

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

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

    SciTech Connect (OSTI)

    Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C.

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    1988-01-01

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

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

    SciTech Connect (OSTI)

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

    2009-04-01

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

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

    SciTech Connect (OSTI)

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

    2015-10-15

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

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

    SciTech Connect (OSTI)

    C. L. Atwood; V. N. Shah; W. J. Galyean

    1999-09-01

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

  9. Pressure suppression containment system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Townsend, Harold E.

    1994-03-15

    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.

  10. Pressure suppression containment system

    DOE Patents [OSTI]

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

    1994-03-15

    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.

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

    SciTech Connect (OSTI)

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

    2000-02-01

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

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

    SciTech Connect (OSTI)

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

    2000-02-01

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

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

    SciTech Connect (OSTI)

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

    2010-12-22

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

  14. A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01

    The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

  15. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

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

  16. Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

    SciTech Connect (OSTI)

    Park, R.J.; Kim, S.B.; Kim, H.D. [Korea Atomic Energy Research Institute, Yuseong, P.O.Box 105, Daejeon, 305-600 (Korea, Republic of)

    2004-07-01

    High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

  17. Radiant vessel auxiliary cooling system

    DOE Patents [OSTI]

    Germer, John H.

    1987-01-01

    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.

  18. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    SciTech Connect (OSTI)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS (1-5) has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code.

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

    SciTech Connect (OSTI)

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

    1986-07-01

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

  20. Pressure and concentration dependences of the autoignition temperature for normal butane + air mixtures in a closed vessel

    SciTech Connect (OSTI)

    Chandraratna, M.R.; Griffiths, J.F. . School of Chemistry)

    1994-12-01

    The condition at which autoignition occurs in lean premixed n-butane + air mixtures over the composition range 0.2%--2.5% n-butane by volume (0.06 < [phi] < 0.66) were investigated experimentally. Total reactant pressure from 0.1 to 0.6 MPa (1--6 atm) were studied in a spherical, stainless-steel, closed vessel (0.5 dm[sup 3]). There is a critical transition from nonignition to ignition, at pressures above 0.1 MPa, as the mixture is enriched in the vicinity of 1% fuel vapor by volume. There is also a region of multiplicity, which exhibits three critical temperatures at a given composition. Chemical analyses show that partially oxygenated components,including many o-heterocyclic compounds, are important products of the lean combustion of butane at temperatures up to 800 K. The critical conditions for autoignition are discussed with regard to industrial ignition hazards, especially in the context of the autoignition temperature of alkanes given by ASTM or BS tests. The differences between the behavior of n-butane and the higher n-alkanes are explained. The experimental results are also used as a basis for testing a reduced kinetic model to represent the oxidation and autoignition of n-butane or other alkanes.

  1. Composition and chemistry of particulates from the Tidd Clean Coal Demonstration Plant pressurized fluidized bed combustor, cyclone, and filter vessel

    SciTech Connect (OSTI)

    Smith, D.H.; Grimm, U.; Haddad, G.

    1995-12-31

    In a Pressurized Fluidized Bed Combustion (PFBC)/cyclone/filter system ground coal and sorbent are injected as pastes into the PFBC bed; the hot gases and entrained fine particles of ash and calcined or reacted sorbent are passed through a cyclone (which removes the larger entrained particles); and the very-fine particles that remain are then filtered out, so that the cleaned hot gas can be sent through a non-ruggedized hot-gas turbine. The 70 MWe Tidd PFBC Demonstration Plant in Brilliant, Ohio was completed in late 1990. The initial design utilized seven strings of primary and secondary cyclones to remove 98% of the particulate matter. However, the Plant also included a pressurized filter vessel, placed between the primary and secondary cyclones of one of the seven strings. Coal and dolomitic limestone (i.e, SO{sub 2} sorbent) of various nominal sizes ranging from 12 to 18 mesh were injected into the combustor operating at about 10 atm pressure and 925{degree}C. The cyclone removed elutriated particles larger than about 0.025 mm, and particles larger than ca. 0.0005 mm were filtered at about 750{degree}C by ceramic candle filters. Thus, the chemical reaction times and temperatures, masses of material, particle-size distributions, and chemical compositions were substantially different for particulates removed from the bed drain, the cyclone drain, and the filter unit. Accordingly, we have measured the particle-size distributions and concentrations of calcium, magnesium, sulfur, silicon, and aluminum for material taken from the three units, and also determined the chemical formulas and predominant crystalline forms of the calcium and magnesium sulfate compounds formed. The latter information is particularly novel for the filter-cake material, from which we isolated the ``new`` compound Mg{sub 2}Ca(SO{sub 4}){sub 3}.

  2. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L.; Lukic, Zarija; Masuda, Koji; Perry, John O.

    2013-05-15

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  3. BIOASSAY VESSEL FAILURE ANALYSIS

    SciTech Connect (OSTI)

    Vormelker, P

    2008-09-22

    Two high-pressure bioassay vessels failed at the Savannah River Site during a microwave heating process for biosample testing. Improper installation of the thermal shield in the first failure caused the vessel to burst during microwave heating. The second vessel failure is attributed to overpressurization during a test run. Vessel failure appeared to initiate in the mold parting line, the thinnest cross-section of the octagonal vessel. No material flaws were found in the vessel that would impair its structural performance. Content weight should be minimized to reduce operating temperature and pressure. Outer vessel life is dependent on actual temperature exposure. Since thermal aging of the vessels can be detrimental to their performance, it was recommended that the vessels be used for a limited number of cycles to be determined by additional testing.

  4. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    DOE Patents [OSTI]

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

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

    SciTech Connect (OSTI)

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

    2009-07-12

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

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

    SciTech Connect (OSTI)

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

    2007-06-30

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

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2015-05-17

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

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

    SciTech Connect (OSTI)

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

    2015-05-17

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

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

    SciTech Connect (OSTI)

    Le Pape, Yann; Huang, Hai

    2016-01-01

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

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

    SciTech Connect (OSTI)

    Pourali, N.; Foroutan, G.

    2015-10-15

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

  11. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

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

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

    SciTech Connect (OSTI)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

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

  13. Pressure suppression system

    DOE Patents [OSTI]

    Gluntz, Douglas M.

    1994-01-01

    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.

  14. Pressure suppression system

    DOE Patents [OSTI]

    Gluntz, D.M.

    1994-10-04

    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. 3 figs.

  15. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28

    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.

  16. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01

    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.

  17. Scanning tunneling microscope assembly, reactor, and system

    DOE Patents [OSTI]

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

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  18. Emergency heat removal system for a nuclear reactor

    DOE Patents [OSTI]

    Dunckel, Thomas L.

    1976-01-01

    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.

  19. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

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

  20. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect (OSTI)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

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

    SciTech Connect (OSTI)

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

    2008-03-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the 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.

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

    SciTech Connect (OSTI)

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

    1981-09-15

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

  3. Evaluation of a dilute chemical decontaminant for pressurized heavy water reactors

    SciTech Connect (OSTI)

    Velmurugan, S.; Narasimhan, S.V.; Mathur, P.K.; Venkateswarlu, K.S. )

    1991-12-01

    In this paper a dilute chemical decontamination formulation based on ethylene diamine tetraacetic acid, oxalic acid, and citric acid is evaluated for its efficacy in removing oxide layers in a pressurized heavy water reactor (PHWR). An ion exchange system that is specifically suited for fission product-dominated contamination in a PHWR is suggested for the reagent regeneration stage of the decontamination process. An attempt has been made to understand the redeposition behavior of various isotopes during the decontamination process. The polarographic method of identifying the species formed in the dissolution process is explained. Electrochemical measurements are employed to follow the course of oxide removal during the dissolution process. Scanning electron micrographs of metal coupons before and after the dissolution process exemplify the involvement of base metal in the formation of a ferrous oxalate layer. Material compatibility tests between the decontaminant and carbon steel, Monel-400, and Zircaloy-2 are reported.

  4. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

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

    SciTech Connect (OSTI)

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

    2011-07-01

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

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

    SciTech Connect (OSTI)

    Pourali, N.; Foroutan, G.

    2015-02-15

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

  7. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    DOE Patents [OSTI]

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  8. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa [Toshiba Corporation (Japan)

    2006-07-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29 m of 1350 MWe ABWR to 26.5 m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage. (authors)

  9. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

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

  10. Control of alkaline stress corrosion cracking in pressurized-water reactor steam generator tubing

    SciTech Connect (OSTI)

    Hwang, I.S. . Dept. of Nuclear Engineering); Park, I.G. . Div. of Materials Science and Engineering)

    1999-06-01

    Outer-diameter stress corrosion cracking (ODSCC) of alloy 600 (UNS N06600) tubings in steam generators of the Kori-1 pressurized-water reactor (PWR) caused an unscheduled outage in 1994. Failure analysis and remedy development studies were undertaken to avoid a recurrence. Destructive examination of a removed tube indicated axial intergranular cracks developed at the top of sludge caused by a boiling crevice geometry. A high ODSCC propagation rate was attributed to high local pH and increased corrosion potential resulting from oxidized copper presumably formed during the maintenance outage and plant heatup. Remedial measures included: (1) crevice neutralization by crevice flushing with boric acid (H[sub 3]BO[sub 3]) and molar ratio control using ammonium chloride (NH[sub 4]Cl), (2) corrosion potential reduction by hydrazine (H[sub 2]NNH[sub 2]) soaking and suppression of oxygen below 20 ppb to avoid copper oxide formation, (3) titanium dioxide (TiO[sub 2]) inhibitor soaking, and (4) temperature reduction of 5 C. Since application of the remedy program, no significant ODSCC has been observed, which clearly demonstrates the benefit of departing from an oxidizing alkaline environment. In addition, the TiO[sub 2] inhibitor appeared to have a positive effect, warranting further examination.

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

    SciTech Connect (OSTI)

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

    1999-02-01

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

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

    SciTech Connect (OSTI)

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

    1999-08-01

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

  13. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

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

  14. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

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

  15. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

    Peinado, Charles O.; Koutz, Stanley L.

    1985-01-01

    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.

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

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

    2016-06-29

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

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

    DOE Patents [OSTI]

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

    2013-09-10

    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.

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

    SciTech Connect (OSTI)

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

    2015-09-01

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

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

    SciTech Connect (OSTI)

    1995-08-01

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

  20. Validation results of the pre-service ultrasonic inspections of the Sizewell B pressurizer and steam generators and reactor coolant pump flywheels

    SciTech Connect (OSTI)

    Conroy, P.J.; Leyland, K.S.

    1995-08-01

    In the UK, concern over the safety issues associated with nuclear power generation resulted in a demand for a public inquiry into the construction and operation of Sizewell ``B``, Britain`s first PWR. This public inquiry was additional to the UK`s normal licensing process. The onus was placed upon the UK utility, CEGB (now Nuclear Electric plc) to provide evidence to the inquiry to support the case that the plant would be constructed and operated to a sufficiently high standard of safety. Part of the evidence to the inquiry (1) relied upon the ability of ultrasonic inspections to verify that the reactor pressure vessel and other safety critical components (collectively known as IoF components), were free from defects that could threaten structural integrity. At that time, the body of evidence showed that although ultrasonic inspection had the potential to satisfy this requirement, it would be necessary to validate the procedures and key operators used in order to provide assurance that they were adequate. Inspection validation therefore became an integral part of the UK PWR nuclear power program.

  1. In-Vessel Torsional Ultrasonic Wave-Based Level Measurement System - Energy

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

    Innovation Portal Advanced Materials Advanced Materials Find More Like This Return to Search In-Vessel Torsional Ultrasonic Wave-Based Level Measurement System Oak Ridge National Laboratory Contact ORNL About This Technology Technology Marketing Summary At Three Mile Island in 1979, a partial meltdown of the core was caused by a sudden, undetected loss of reactor coolant water. In the past, a reactor's high temperature and pressure environment has complicated the implementation of level

  2. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

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

  3. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    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 (top), then places it ...

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

    SciTech Connect (OSTI)

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

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  6. Operating experience feedback report -- Pressure locking and thermal binding of gate valves. Commercial power reactors: Volume 9

    SciTech Connect (OSTI)

    Hsu, C.

    1993-03-01

    The potential for valve inoperability caused by pressure locking and thermal binding has been known for many years in the nuclear industry. Pressure locking or thermal binding is a common-mode failure mechanism that can prevent a gate valve from opening, and could render redundant trains of safety systems or multiple safety systems inoperable. In spite of numerous generic communications issued in the past by the Nuclear Regulatory Commission (NRC) and industry, pressure locking and thermal binding continues to occur to gate valves installed in safety-related systems of both boding water reactors (BWRs) and pressurized water reactors (PWRs). The generic communications to date have not led to effective industry action to fully identify, evaluate, and correct the problem. This report provides a review of operating events involving these failure mechanisms. As a result of this review this report: (1) identifies conditions when the failure mechanisms have occurred, (2) identifies the spectrum of safety systems that have been subjected to the failure mechanisms, and (3) identifies conditions that may introduce the failure mechanisms under both normal and accident conditions. On the basis of the evaluation of the operating events, the Office for Analysis and Evaluation of Operational Data (AEOD) of the NRC concludes that the binding problems with gate valves are an important safety issue that needs priority NRC and industry attention. This report also provides AEOD`s recommendation for actions to effectively prevent the occurrence of valve binding failures.

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

    SciTech Connect (OSTI)

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

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

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    George, Nathan Michael; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Maldonado, Ivan

    2014-09-29

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

  9. Dissolver vessel bottom assembly

    DOE Patents [OSTI]

    Kilian, Douglas C.

    1976-01-01

    An improved bottom assembly is provided for a nuclear reactor fuel reprocessing dissolver vessel wherein fuel elements are dissolved as the initial step in recovering fissile material from spent fuel rods. A shock-absorbing crash plate with a convex upper surface is disposed at the bottom of the dissolver vessel so as to provide an annular space between the crash plate and the dissolver vessel wall. A sparging ring is disposed within the annular space to enable a fluid discharged from the sparging ring to agitate the solids which deposit on the bottom of the dissolver vessel and accumulate in the annular space. An inlet tangential to the annular space permits a fluid pumped into the annular space through the inlet to flush these solids from the dissolver vessel through tangential outlets oppositely facing the inlet. The sparging ring is protected against damage from the impact of fuel elements being charged to the dissolver vessel by making the crash plate of such a diameter that the width of the annular space between the crash plate and the vessel wall is less than the diameter of the fuel elements.

  10. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

    Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

    1987-01-01

    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.

  11. Start-up control system and vessel for LMFBR

    DOE Patents [OSTI]

    Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

    1987-01-01

    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.

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

    SciTech Connect (OSTI)

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

    2010-10-01

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

  13. Research Update: Atmospheric pressure spatial atomic layer deposition of ZnO thin films: Reactors, doping, and devices

    SciTech Connect (OSTI)

    Hoye, Robert L. Z. E-mail: jld35@cam.ac.uk; MacManus-Driscoll, Judith L. E-mail: jld35@cam.ac.uk; Muoz-Rojas, David; Nelson, Shelby F.; Illiberi, Andrea; Poodt, Paul

    2015-04-01

    Atmospheric pressure spatial atomic layer deposition (AP-SALD) has recently emerged as an appealing technique for rapidly producing high quality oxides. Here, we focus on the use of AP-SALD to deposit functional ZnO thin films, particularly on the reactors used, the film properties, and the dopants that have been studied. We highlight how these films are advantageous for the performance of solar cells, organometal halide perovskite light emitting diodes, and thin-film transistors. Future AP-SALD technology will enable the commercial processing of thin films over large areas on a sheet-to-sheet and roll-to-roll basis, with new reactor designs emerging for flexible plastic and paper electronics.

  14. RSMASS: A simple model for estimating reactor and shield masses

    SciTech Connect (OSTI)

    Marshall, A.C.; Aragon, J.; Gallup, D.

    1987-01-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, thermal/hydraulic limits, or fuel damage limits, whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should be applicable to a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations.

  15. Particle Imaging Velocimetry Technique Development for Laboratory Measurement of Fracture Flow Inside a Pressure Vessel Using Neutron Imaging

    SciTech Connect (OSTI)

    Polsky, Yarom; Bingham, Philip R; Bilheux, Hassina Z; Carmichael, Justin R

    2015-01-01

    This paper will describe recent progress made in developing neutron imaging based particle imaging velocimetry techniques for visualizing and quantifying flow structure through a high pressure flow cell with high temperature capability (up to 350 degrees C). This experimental capability has great potential for improving the understanding of flow through fractured systems in applications such as enhanced geothermal systems (EGS). For example, flow structure measurement can be used to develop and validate single phase flow models used for simulation, experimentally identify critical transition regions and their dependence on fracture features such as surface roughness, and study multiphase fluid behavior within fractured systems. The developed method involves the controlled injection of a high contrast fluid into a water flow stream to produce droplets that can be tracked using neutron radiography. A description of the experimental setup will be provided along with an overview of the algorithms used to automatically track droplets and relate them to the velocity gradient in the flow stream. Experimental results will be reported along with volume of fluids based simulation techniques used to model observed flow.

  16. Nuclear reactor building

    DOE Patents [OSTI]

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

    1994-04-05

    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.

  17. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    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.

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

    SciTech Connect (OSTI)

    Lata

    1996-09-26

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

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

    SciTech Connect (OSTI)

    Lancaster, D.

    1992-10-26

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

  20. High flux reactor

    DOE Patents [OSTI]

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

    1988-01-01

    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.

  1. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

    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.

  2. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect (OSTI)

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  3. An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

    SciTech Connect (OSTI)

    Oßwald, Patrick; Köhler, Markus

    2015-10-15

    A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimental data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.

  4. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1987-01-01

    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.

  5. Boiling water neutronic reactor incorporating a process inherent safety design

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19

    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.

  6. FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

    SciTech Connect (OSTI)

    McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.; Morgan, W.C.; Suski, A.E.

    1988-05-01

    This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15 refs., 6 figs., 5 tabs.

  7. Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

    SciTech Connect (OSTI)

    Schreiber, Daniel K.; Olszta, Matthew J.; Saxey, David W.; Kruska, Karen; Moore, K. L.; Lozano-Perez, Sergio; Bruemmer, Stephen M.

    2013-06-01

    High-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325 C simulated pressurized water reactor (PWR) primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

  8. Containment of explosions in spherical vessels

    SciTech Connect (OSTI)

    Duffey, T.A.; Greene, J.M. ); Baker, W.E. . Dept. of Mechanical Engineering); Lewis, B.B. )

    1992-01-01

    A correlation of the experimentally recorded dynamic response of a spherical containment vessel with theoretical finite element calculations is presented. Three experiments were performed on the 6-ft-diameter steel vessel using centrally located 12-lb. and 40-lb. high explosive charges. Pressure-time loading on the inner wall of the vessel was recorded for each test using pressure transducers. Resulting dynamic response of the vessel was recorded for each test using strain gages mounted at selected locations on the outer surface of the vessel. Response of the vessel was primarily elastic. A finite element model of the vessel was run using DYNA3D, a dynamic structural analysis code. Pressure loading for the finite element model was based on results from a one-dimensional reactive hydrodynamics code. Correlations between experiments and analysis were generally good for the tests for frequency and strain magnitude at most locations. Comparisons of experimental and calculated pressure-time histories were less satisfactory.

  9. Containment of explosions in spherical vessels

    SciTech Connect (OSTI)

    Duffey, T.A.; Greene, J.M.; Baker, W.E.; Lewis, B.B.

    1992-12-31

    A correlation of the experimentally recorded dynamic response of a spherical containment vessel with theoretical finite element calculations is presented. Three experiments were performed on the 6-ft-diameter steel vessel using centrally located 12-lb. and 40-lb. high explosive charges. Pressure-time loading on the inner wall of the vessel was recorded for each test using pressure transducers. Resulting dynamic response of the vessel was recorded for each test using strain gages mounted at selected locations on the outer surface of the vessel. Response of the vessel was primarily elastic. A finite element model of the vessel was run using DYNA3D, a dynamic structural analysis code. Pressure loading for the finite element model was based on results from a one-dimensional reactive hydrodynamics code. Correlations between experiments and analysis were generally good for the tests for frequency and strain magnitude at most locations. Comparisons of experimental and calculated pressure-time histories were less satisfactory.

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

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

    SciTech Connect (OSTI)

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

    2003-11-12

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

  12. Support arrangement for core modules of nuclear reactors

    DOE Patents [OSTI]

    Bollinger, Lawrence R.

    1987-01-01

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  13. Support arrangements for core modules of nuclear reactors. [PWR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  14. Thermo-hydraulic Simulation of Pressurizer in Transient Cases

    SciTech Connect (OSTI)

    Ardeshir, A.T.; Nematollahi, M.; Sepanloo, K.; Daneshvari, F.

    2004-07-01

    This paper describes a simulation of the pressure adjustment in the primary loop of Pressurized Water Reactors (PWR). A mathematical model is developed for the thermo-hydraulic behavior of pressurizer in transient cases (surge in or out) on the basis of concept of conservation of mass and energy in two phases. No restrictive assumptions have been made. A comparison with RELAP5/Mod3.2 data indicates good overall agreement. The model can be used as a good design verification tool for pressurizer vessels and associated pressure control devices. (authors)

  15. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    SciTech Connect (OSTI)

    Rahman, Fariz Abdul; Lee, John C. [University of Michigan, Ann Arbor, MI (United States); Franceschini, Fausto; Wenner, Michael [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and

  16. RSMASS: A preliminary reactor/shield mass model for SDI applications

    SciTech Connect (OSTI)

    Marshall, A.C.

    1986-08-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, specific power limits, or fuel burnup limits - whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should provide good mass estimates for a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations.

  17. Inservice leak testing of primary pressure isolation valves. Final report

    SciTech Connect (OSTI)

    Livingston, R.A.

    1983-02-01

    This report discusses the inservice leak testing of primary pressure isolation valves in commercial power reactors which was investigated to identify problems with current test procedures and requirements. Nine utilities were surveyed to gather information which is presented in this report. An analysis of the survey information was performed, resulting in recommended changes to improve valve leak testing requirements currently invoked by Section XI of the ASME Boiler and Pressure Vessel Code, Plant Technical Specifications, and Regulatory Guides addressing this subject.

  18. Inservice leak testing of primary pressure isolation valves

    SciTech Connect (OSTI)

    Livingston, R.A.

    1983-02-01

    This report discusses the inservice leak testing of primary pressure isolation valves in commercial power reactors which was investigated to identify problems with current test procedures and requirements. Nine utilities were surveyed to gather information which is presented in this report. An analysis of the survey information was performed, resulting in recommended changes to improve valve leak testing requirements currently invoked by Section XI of the ASME Boiler and Pressure Vessel Code, Plant Technical Specifications, and Regulatory Guides addressing this subject.

  19. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect (OSTI)

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  20. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle...

    Office of Scientific and Technical Information (OSTI)

    a Pressurized Water Reactor Citation Details In-Document Search Title: Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor You ...

  1. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle...

    Office of Scientific and Technical Information (OSTI)

    a Pressurized Water Reactor Citation Details In-Document Search Title: Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor ...

  2. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  3. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOE Patents [OSTI]

    Moore, W.T.

    1958-09-01

    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.

  4. Combustion synthesis continuous flow reactor

    DOE Patents [OSTI]

    Maupin, Gary D.; Chick, Lawrence A.; Kurosky, Randal P.

    1998-01-01

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

  5. Combustion synthesis continuous flow reactor

    DOE Patents [OSTI]

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  6. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    SciTech Connect (OSTI)

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  7. Steam Gasification Rates of Three Bituminous Coal Chars in an Entrained-Flow Reactor at Pressurized Conditions

    SciTech Connect (OSTI)

    Lewis, Aaron D.; Holland, Troy M.; Marchant, Nathaniel R.; Fletcher, Emmett G.; Henley, Daniel J.; Fuller, Eric G.; Fletcher, Thomas H.

    2015-02-26

    Three bituminous coal chars (Illinois #6, Utah Skyline, and Pittsburgh #8) were gasified separately at total pressures of 10 and 15 atm in an entrained-flow reactor using gas temperatures up to 1830 K and particle residence times <240 ms. The experiments were performed at conditions where the majority of particle mass release was due to H2O gasification, although select experiments were performed at conditions where significant mass release was due to gasification by both H2O and CO2. The measured coal data we recorded were fit to three char gasification models including a simple first-order global model, as well as the CCKNand CCK models that stem from the CBK model. The optimal kinetic parameters for each of the three models are reported, and the steam reactivity of the coal chars at the studied conditions is as follows: Pittsburgh #8 > Utah Skyline > Illinois #6.

  8. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect (OSTI)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  9. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

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

  10. Pressure regulator

    DOE Patents [OSTI]

    Ebeling, Jr., Robert W.; Weaver, Robert B.

    1979-01-01

    The pressure within a pressurized flow reactor operated under harsh environmental conditions is controlled by establishing and maintaining a fluidized bed of uniformly sized granular material of selected density by passing the gas from the reactor upwardly therethrough at a rate sufficient to fluidize the bed and varying the height of the bed by adding granular material thereto or removing granular material therefrom to adjust the backpressure on the flow reactor.

  11. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

    Office of Environmental Management (EM)

    Reactor Vessel Manufacturing Within a Factory Environment - Volume 1 This study focused on the learning process for the factory built components of the Integrated Reactor ...

  12. LPG storage vessel cracking experience

    SciTech Connect (OSTI)

    Cantwell, J.E. )

    1988-10-01

    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.

  13. LPG storage vessel cracking experience

    SciTech Connect (OSTI)

    Cantwell, J.E.

    1988-01-01

    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.

  14. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    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.

  15. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    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.

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-04-20

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3 and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistrymore » in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from VOCs, NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach two orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s−1) ten-fold OH suppression is observed at OHRext ∼ 100 s−1, which is similar or lower than used in many laboratory studies. This finding may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over an order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross-sections in the model are within ± 25% for OH exposure and within ± 60% for other parameters. These uncertainties are small relative to the dynamic range of outputs. Uncertainty

  17. Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

    SciTech Connect (OSTI)

    Nad, Shreya; Gu, Yajun; Asmussen, Jes

    2015-07-15

    The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficiencies (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.

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

    SciTech Connect (OSTI)

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

    1997-03-01

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

  19. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    SciTech Connect (OSTI)

    Syarifah, Ratna Dewi Suud, Zaki

    2015-09-30

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  20. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1962-08-14

    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)

  1. Liquid level, void fraction, and superheated steam sensor for nuclear reactor cores

    DOE Patents [OSTI]

    Tokarz, Richard D. (West Richland, WA)

    1983-01-01

    An apparatus for detecting nominal phase conditions of coolant in a reactor vessel comprising one or more lengths of tubing each leading from a location being monitored to a closed outer end exterior of the vessel. Temperature is sensed at the open end of each length of tubing. Pressure within the tubing is also sensed. Both measurements are directed to an analyzer which compares the measured temperature to the known saturated temperature of the coolant at the measured pressure. In this manner, the nominal phase conditions of the coolant are constantly monitored.

  2. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    SciTech Connect (OSTI)

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  3. 6151 Pressure Systems

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

    For design, fabrication, testing, repair, modification and inspection are based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section ...

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Peng, Z.; Day, D. A.; Stark, H.; Li, R.; Lee-Taylor, J.; Palm, B. B.; Brune, W. H.; Jimenez, J. L.

    2015-11-20

    Oxidation flow reactors (OFRs) using OH produced from low-pressure Hg lamps at 254 nm (OFR254) or both 185 and 254 nm (OFR185) are commonly used in atmospheric chemistry and other fields. OFR254 requires the addition of externally formed O3 since OH is formed from O3 photolysis, while OFR185 does not since O2 can be photolyzed to produce O3, and OH can also be formed from H2O photolysis. In this study, we use a plug-flow kinetic model to investigate OFR properties under a very wide range of conditions applicable to both field and laboratory studies. We show that the radical chemistrymore » in OFRs can be characterized as a function of UV light intensity, H2O concentration, and total external OH reactivity (OHRext, e.g., from volatile organic compounds (VOCs), NOx, and SO2). OH exposure is decreased by added external OH reactivity. OFR185 is especially sensitive to this effect at low UV intensity due to low primary OH production. OFR254 can be more resilient against OH suppression at high injected O3 (e.g., 70 ppm), as a larger primary OH source from O3, as well as enhanced recycling of HO2 to OH, make external perturbations to the radical chemistry less significant. However if the external OH reactivity in OFR254 is much larger than OH reactivity from injected O3, OH suppression can reach 2 orders of magnitude. For a typical input of 7 ppm O3 (OHRO3 = 10 s−1), 10-fold OH suppression is observed at OHRext ~ 100 s−1, which is similar or lower than used in many laboratory studies. The range of modeled OH suppression for literature experiments is consistent with the measured values except for those with isoprene. The finding on OH suppression may have important implications for the interpretation of past laboratory studies, as applying OHexp measurements acquired under different conditions could lead to over a 1-order-of-magnitude error in the estimated OHexp. The uncertainties of key model outputs due to uncertainty in all rate constants and absorption cross

  5. Alternative Passive Decay-Heat Systems for the Advanced High-Temperature Reactor

    SciTech Connect (OSTI)

    Forsberg, Charles W.

    2006-07-01

    The Advanced High-Temperature Reactor (AHTR) is a low-pressure, liquid-salt-cooled high-temperature reactor for the production of electricity and hydrogen. The high-temperature (950 deg C) variant is defined as the liquid-salt-cooled very high-temperature reactor (LS-VHTR). The AHTR has the same safety goals and uses the same graphite-matrix coated particle fuel as do modular high-temperature gas-cooled reactors. However, the large AHTR power output [2400 to 4000 MW(t)] implies the need for a different type of passive decay-heat removal system. Because the AHTR is a low-pressure, liquid-cooled reactor like sodium-cooled reactors, similar types of decay-heat-removal systems can be used. Three classes of passive decay heat removal systems have been identified: the reactor vessel auxiliary cooling system which is similar to that proposed for the General Electric S-PRISM sodium-cooled fast reactor; the direct reactor auxiliary cooling system, which is similar to that used in the Experimental Breeder Reactor-II; and a new pool reactor auxiliary cooling system. These options are described and compared. (author)

  6. Advanced Nuclear Reactors | Department of Energy

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

    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

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

    SciTech Connect (OSTI)

    Morfin, Franck; Piccolo, Laurent

    2013-09-15

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

  8. Microsoft Word - Errata for the Pressure and Vacuum Systems Safety Supplement 3-15

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

    PRESSURE VESSEL REGISTRATION FORM PS-4 Pressure System Number: Date: Pressure System Name: Pressure Vessel Number: P&ID Number: Pressure Vessel Description: MAWP/Design Pressure: Design Temperature: Operating Pressure: Operating Temperature: Code: Code Year: System Fluid: Fluid Category: Fluid State: VESSEL DATA ASME Stamp Type ___U Stamp ____UM Stamp ___Other (specify) Vessel Type: __Air Tank __Water Tank __Non-Flam Gas Tank __Flam Gas Tank __Other (specify) Vessel Manufacturer National

  9. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOE Patents [OSTI]

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

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

    SciTech Connect (OSTI)

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

    1994-01-01

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

  11. Vertical Pretreatment Reactor System (Poster), NREL (National...

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

    Vertical Pretreatment Reactor System Two-vessel system for primary and secondary ... moves by gravity from top to bottom of reactor in plug-fl ow fashion * Residence time is ...

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

    SciTech Connect (OSTI)

    Not Available

    1980-05-01

    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;

  13. 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 (OSTI)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    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

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

    SciTech Connect (OSTI)

    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.

    1982-07-01

    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.

  15. Draft ASME Boiler and Pressure Vessel Code Section III, Division 5, Section HB, Subsection B, Code Case for Alloy 617 and Background Documentation

    SciTech Connect (OSTI)

    Wright, Julie Knibloe

    2015-08-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger for the very high temperature reactor. To evaluate the behavior of this material in the expected service conditions, strain controlled cyclic tests that include long hold times up to 240 minutes at maximum tensile strain were conducted at 850°C. In terms of the total number of cycles to failure, the fatigue resistance decreased when a hold time was added at peak tensile strain. Increases in the tensile hold duration degraded the creep fatigue resistance, at least to the investigated strain controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. The creep fatigue deformation mode is considered relative to the lack of saturation, or continually decreasing number of cycles to failure with increasing hold times. Additionally, preliminary values from the 850°C creep fatigue data are calculated for the creep fatigue damage diagram and have higher values of creep damage than those from tests at 950°C.

  16. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    SciTech Connect (OSTI)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  17. Structural Analysis of the NCSX Vacuum Vessel

    SciTech Connect (OSTI)

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-09-28

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered.

  18. PR-EDB: Power Reactor Embrittlement Database - Version 3

    SciTech Connect (OSTI)

    Wang, Jy-An John; Subramani, Ranjit

    2008-03-01

    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

  19. Treating exhaust gas from a pressurized fluidized bed reaction system

    DOE Patents [OSTI]

    Isaksson, J.; Koskinen, J.

    1995-08-22

    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.

  20. Treating exhaust gas from a pressurized fluidized bed reaction system

    DOE Patents [OSTI]

    Isaksson, Juhani; Koskinen, Jari

    1995-01-01

    Hot gases from a pressurized fluidized bed reactor system are purified. Under superatmospheric pressure conditions hot exhaust gases are passed through a particle separator, forming a flitrate 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 superatmospheric 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).

  1. Coal gasification vessel

    DOE Patents [OSTI]

    Loo, Billy W.

    1982-01-01

    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).

  2. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    SciTech Connect (OSTI)

    Rohanda, Anis; Waris, Abdul

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  3. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOE Patents [OSTI]

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    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.

  4. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

    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.

  5. naval reactors | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    on Energy and Water Development, visited the Naval Reactors Facility (NRF) at the... ... propulsion plants use a pressurized-water reactor design that has two basic systems: ...

  6. EDS V25 containment vessel explosive qualification test report.

    SciTech Connect (OSTI)

    Rudolphi, John Joseph

    2012-04-01

    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.

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

    SciTech Connect (OSTI)

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

    2015-08-31

    Gas discharge plasmas used for thinfilm deposition by plasma-enhanced chemical vapor deposition (PECVD) must be devoid of contaminants, like dust or active species which disturb the intended chemical reaction. In atmospheric pressure plasma systems employing an inert gas, the main source of such contamination is the residual air inside the system. To enable the construction of an atmospheric pressure plasma (APP) system with minimal contamination, we have carried out fluid dynamic simulation of the APP chamber into which an inert gas is injected at different mass flow rates. On the basis of the simulation results, we have designed and built a simple, scaled APP system, which is capable of holding a 100 mm substrate wafer, so that the presence of air (contamination) in the APP chamber is minimized with as low a flow rate of argon as possible. This is examined systematically by examining optical emission from the plasma as a function of inert gas flow rate. It is found that optical emission from the plasma shows the presence of atmospheric air, if the inlet argon flow rate is lowered below 300 sccm. That there is minimal contamination of the APP reactor built here, was verified by conducting an atmospheric pressure PECVD process under acetylene flow, combined with argon flow at 100 sccm and 500 sccm. The deposition of a polymer coating is confirmed by infrared spectroscopy. X-ray photoelectron spectroscopy shows that the polymer coating contains only 5% of oxygen, which is comparable to the oxygen content in polymer deposits obtained in low-pressure PECVD systems.

  8. Defective fuel rod detection in operating pressurized water reactors during periods of continuously decreasing fuel rod integrity levels

    SciTech Connect (OSTI)

    Zanker, H. )

    1989-09-01

    Periods of continuously decreasing levels of fuel rod integrity due to debris-induced cladding damage, vibration-induced fretting wear of the cladding, etc. cause difficulties in the assessment of fuel rod performance from coolant activity data. The calculational models currently in use for this purpose in nuclear power plants are not sufficiently capable of indicating cases in which they are invalid. This can mislead reactor operators by misinterpretation of the coolant activity data, especially in situations where fast reactions are necessary. A quick test of validity is suggested to check the applicability of the currently available calculational models for estimating the number and average size of fuel rod defects. This paper describes how to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation.

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

    SciTech Connect (OSTI)

    Tusheva, P.; Schaefer, F.; Kliem, S.

    2012-07-01

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

  10. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

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

  11. Static-stress analysis of dual-axis safety vessel

    SciTech Connect (OSTI)

    Bultman, D.H.

    1992-11-01

    An 8-ft-diameter safety vessel, made of HSLA-100 steel, is evaluated to determine its ability to contain the quasi-static residual pressure from a high-explosive (HE) blast. The safety vessel is designed for use with the Dual-Axis Radiographic Hydrotest (DARHT) facility being developed at Los Alamos National Laboratory. A smaller confinement vessel fits inside the safety vessel and contains the actual explosion, and the safety vessel functions as a second layer of containment in the unlikely case of a confinement vessel leak. The safety vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, and the Welding Research Council Bulletin, WRC107. Combined stresses that result from internal pressure and external loads on nozzles are calculated and compared to the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzle components are adequately designed for a static pressure of 830 psi, plus the maximum expected external loads. Shell stresses at the shellto-nozzle interface, produced from external loads on the nozzles, were less than 700 psi. The maximum combined stress resulting from the internal pressure plus external loads was 17,384 psi, which is significantly less than the allowable stress of 42,375 psi for HSLA-100 steel.

  12. Assessment of Homogeneous Thorium/Uranium Fuel for Pressurized...

    Office of Scientific and Technical Information (OSTI)

    Fuel for Pressurized Water Reactors Citation Details In-Document Search Title: Assessment of Homogeneous ThoriumUranium Fuel for Pressurized Water Reactors The homogeneous ...

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

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

    Energy of Large Stationary Hydrogen/CNG/HCNG Storage Vessels R&D of Large Stationary Hydrogen/CNG/HCNG Storage Vessels These slides were presented at the International Hydrogen Fuel and Pressure Vessel Forum on September 27 - 29, 2010, in Beijing, China. ihfpv_zheng2.pdf (1.54 MB) More Documents & Publications Forum Agenda: International Hydrogen Fuel and Pressure Vessel Forum Bonfire Tests of High Pressure Hydrogen Storage Tanks Status and Progress in Research, Development and

  14. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05

    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

  15. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2011-10-13

    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

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

    SciTech Connect (OSTI)

    Kck, Eva-Maria; Kogler, Michaela; Pramsoler, Reinhold; Kltzer, Bernhard; Penner, Simon

    2014-08-15

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

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

    DOE Patents [OSTI]

    Holmes, Michael Jerome; Ohrn, Theodore R.; Chen, Christopher Ming-Poh

    2010-02-09

    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.

  18. Cryogenic Pressure Vessels: Progress and Plans

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

    Salvador Aceves, Gene Berry, Francisco Espinosa, Ibo Matthews, Guillaume Petitpas, Tim Ross, Ray Smith, Vernon Switzer Lawrence Livermore National Laboratory February 15, 2011 This ...

  19. An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

    SciTech Connect (OSTI)

    Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

    2008-01-01

    Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, “Inservice Inspection of Nuclear Power Plant Components,” with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper.

  20. Catalytic reactor

    DOE Patents [OSTI]

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

    2009-03-10

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

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    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.

  2. A resting bottom sodium cooled fast reactor

    SciTech Connect (OSTI)

    Costes, D.

    2012-07-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  3. Static-stress analysis of dual-axis confinement vessel

    SciTech Connect (OSTI)

    Bultman, D.H.

    1992-11-01

    This study evaluates the static-pressure containment capability of a 6-ft-diameter, spherical vessel, made of HSLA-100 steel, to be used for high-explosive (HE) containment. The confinement vessel is designed for use with the Dual-Axis Radiographic Hydrotest Facility (DARHT) being developed at Los Alamos National Laboratory. Two sets of openings in the vessel are covered with x-ray transparent covers to allow radiographic imaging of an explosion as it occurs inside the vessel. The confinement vessel is analyzed as a pressure vessel based on the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, and the Welding Research Council Bulletin, WRC-107. Combined stresses resulting from internal pressure and external loads on nozzles are calculated and compared with the allowable stresses for HSLA-100 steel. Results confirm that the shell and nozzles of the confinement vessel are adequately designed to safely contain the maximum residual pressure of 1675 psi that would result from an HE charge of 24.2 kg detonated in a vacuum. Shell stresses at the shell-to-nozzle interface, produced from external loads on the nozzles, were less than 400 psi. The maximum combined stress resulting from the internal pressure plus external loads was 16,070 psi, which is less than half the allowable stress of 42,375 psi for HSLA-100 steel.

  4. Relief device for a vacuum vessel

    DOE Patents [OSTI]

    Fast, Ronald W.

    1987-04-28

    A pressure relief device 5 for a vessel having redundant pressure relief capabilities. An annular plate 12 overlies a surface 11 which has an aperature to the vessel. A seal is formed between the surface 11 and annular plate 12. A solid plate 13 overlies the annular plate 12. A seal is formed between the solid plate 13 and annular plate 12. The relief device 5 will open at a first predetermined pressure by lifting the solid plate 13. In the event the seal between solid plate 13 and annular plate 12 should stick the relief device 5 will open at a second slightly higher, predetermined pressure by lifting the annular plate 12 and solid plate 13 together. Hinging means 6 are provided to reclose the pressure relief device 5 when conditions return to normal.

  5. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  6. PRESSURE SYSTEM CONTROL

    DOE Patents [OSTI]

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

    1961-06-20

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

  7. Water-Gas-Shift Membrane Reactor for High-Pressure Hydrogen Production. A comprehensive project report (FY2010 - FY2012)

    SciTech Connect (OSTI)

    Klaehn, John; Peterson, Eric; Orme, Christopher; Bhandari, Dhaval; Miller, Scott; Ku, Anthony; Polishchuk, Kimberly; Narang, Kristi; Singh, Surinder; Wei, Wei; Shisler, Roger; Wickersham, Paul; McEvoy, Kevin; Alberts, William; Howson, Paul; Barton, Thomas; Sethi, Vijay

    2013-01-01

    Idaho National Laboratory (INL), GE Global Research (GEGR), and Western Research Institute (WRI) have successfully produced hydrogen-selective membranes for water-gas-shift (WGS) modules that enable high-pressure hydrogen product streams. Several high performance (HP) polymer membranes were investigated for their gas separation performance under simulated (mixed gas) and actual syngas conditions. To enable optimal module performance, membranes with high hydrogen (H2) selectivity, permeance, and stability under WGS conditions are required. The team determined that the VTEC PI 80-051 and VTEC PI 1388 (polyimide from Richard Blaine International, Inc.) are prime candidates for the H2 gas separations at operating temperatures (~200C). VTEC PI 80-051 was thoroughly analyzed for its H2 separations under syngas processing conditions using more-complex membrane configurations, such as tube modules and hollow fibers. These membrane formats have demonstrated that the selected VTEC membrane is capable of providing highly selective H2/CO2 separation (? = 7-9) and H2/CO separation (? = 40-80) in humidified syngas streams. In addition, the VTEC polymer membranes are resilient within the syngas environment (WRI coal gasification) at 200C for over 1000 hours. The information within this report conveys current developments of VTEC PI 80-051 as an effective H2 gas separations membrane for high-temperature syngas streams.

  8. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  9. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    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.

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

    SciTech Connect (OSTI)

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

    1994-05-01

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

  11. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  12. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  13. Flooding Experiments and Modeling for Improved Reactor Safety

    SciTech Connect (OSTI)

    Solmos, M., Hogan, K.J., VIerow, K.

    2008-09-14

    Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.

  14. GAS COOLED NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  15. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

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

  16. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

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

  17. Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology

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

    | Department of Energy Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology Inexpensive Delivery of Compressed Hydrogen with Advanced Vessel Technology Presentation on inexpensive delivery of compressed hydrogen with advanced vessel technology. wkshp_storage_berry.pdf (367.52 KB) More Documents & Publications President's Hydrogen Fuel Initiative Overview of FreedomCAR & Fuels Partnership/DOE Delivery Program High-Pressure Tube Trailers and Tanks

  18. Using the OECD/NRC Pressurized Water Reactor Main Steam Line Break Benchmark to Study Current Numerical and Computational Issues of Coupled Calculations

    SciTech Connect (OSTI)

    Ivanov, Kostadin N.; Todorova, Nadejda K.; Sartori, Enrico

    2003-05-15

    Incorporating full three-dimensional (3-D) models of the reactor core into system transient codes allows for a 'best-estimate' calculation of interactions between the core behavior and plant dynamics. Recent progress in computer technology has made the development of coupled thermal-hydraulic (T-H) and neutron kinetics code systems feasible. Considerable efforts have been made in various countries and organizations in this direction. Appropriate benchmarks need to be developed that will permit testing of two particular aspects. One is to verify the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. The second is to test fully the neutronics/T-H coupling. One such benchmark is the Pressurized Water Reactor Main Steam Line Break (MSLB) Benchmark problem. It was sponsored by the Organization for Economic Cooperation and Development, U.S. Nuclear Regulatory Commission, and The Pennsylvania State University. The benchmark problem uses a 3-D neutronics core model that is based on real plant design and operational data for the Three Mile Island Unit 1 nuclear power plant. The purpose of this benchmark is threefold: to verify the capability of system codes for analyzing complex transients with coupled core-plant interactions; to test fully the 3-D neutronics/T-H coupling; and to evaluate discrepancies among the predictions of coupled codes in best-estimate transient simulations. The purposes of the benchmark are met through the application of three exercises: a point kinetics plant simulation (exercise 1), a coupled 3-D neutronics/core T-H evaluation of core response (exercise 2), and a best-estimate coupled core-plant transient model (exercise 3).In this paper we present the three exercises of the MSLB benchmark, and we summarize the findings of the participants with regard to the current numerical and computational issues of coupled calculations. In addition, this paper reviews in some detail the sensitivity studies on

  19. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    SciTech Connect (OSTI)

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  20. The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

    SciTech Connect (OSTI)

    Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

    2006-11-01

    Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplamenting volumetric metods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural intefrity of reactor components. As piping and pressure vessel compoents in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conductrd a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to

  1. Control system for a small fission reactor

    DOE Patents [OSTI]

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  2. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  3. Containment pressurization and burning of combustible gases in a large, dry PWR containment during a station blackout sequence

    SciTech Connect (OSTI)

    Lee, M.; Fan, C.T. (National Tsing-Hua Univ., Dept. of Nuclear Engineering, Hsinchu (TW))

    1992-07-01

    In this paper, responses of a large, dry pressurized water reactor (PWR) containment in a station blackout sequence are analyzed with the CONTAIN, MARCH3, and MAAP codes. Results show that the predicted containment responses in a station blackout sequence of these three codes are substantially different. Among these predictions, the MAAP code predicts the highest containment pressure because of the large amount of water made available to quench the debris upon vessel failure. The gradual water boiloff by debris pressurizes the containment. The combustible gas burning models in these codes are briefly described and compared.

  4. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

  5. Automatic safety rod for reactors

    DOE Patents [OSTI]

    Germer, John H.

    1988-01-01

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

  6. WATER BOILER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

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

  8. Optimized Battery-Type Reactor Primary System Design Utilizing Lead

    SciTech Connect (OSTI)

    Yu, Yong H.; Son, Hyoung M.; Lee, Il S.; Suh, Kune Y.

    2006-07-01

    A number of small and medium size reactors are being developed worldwide as well as large electricity generation reactors for co-generation, district heating or desalination. The Seoul National University has started to develop 23 MWth BORIS (Battery Optimized Reactor Integral System) as a multi-purpose reactor. BORIS is an integral-type optimized fast reactor with an ultra long life core. BORIS is being designed to meet the Generation IV nuclear energy system goals of sustainability, safety, reliability and economics. Major features of BORIS include 20 consecutive years of operation without refueling; elimination of an intermediate heat transport loop and main coolant pump; open core without individual subassemblies; inherent negative reactivity feedback; and inherent load following capability. Its one mission is to provide incremental electricity generation to match the needs of developing nations and especially remote communities without major electrical grid connections. BORIS consists of a reactor module, heat exchanger, coolant module, guard vessel, reactor vessel auxiliary cooling system (RVACS), secondary system, containment and the seismic isolation. BORIS is designed to generate 10 MWe with the resulting thermal efficiency of 45 %. BORIS uses lead as the primary system coolant because of the inherent safety of the material. BORIS is coupled with a supercritical carbon dioxide Brayton cycle as the secondary system to gain a high cycle efficiency in the range of 45 %. The reference core consists of 757 fuel rods without assembly with an active core height of 0.8 m. The BORIS core consists of single enrichment zone composed of a Pu-MA (minor actinides)-U-N fuel and a ferritic-martensitic stainless steel clad. This study is intended to set up appropriate reactor vessel geometry by performing thermal hydraulic analysis on RVACS using computational fluid dynamics codes; to examine the liquid metal coolant behavior along the subchannels; to find out whether the

  9. Tow Vessel | Open Energy Information

    Open Energy Info (EERE)

    Tow Vessel Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleTowVessel&oldid596390" Feedback Contact needs updating Image needs updating...

  10. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

    Sutton, Jr., Harry G.

    1987-01-01

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

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

    SciTech Connect (OSTI)

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

    2012-09-01

    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

  12. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-07-05

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

  13. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, Philippe

    1994-01-01

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

  14. LANL Robotic Vessel Scanning

    SciTech Connect (OSTI)

    Webber, Nels W.

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  15. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    SciTech Connect (OSTI)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  16. CONTROL MEANS FOR A NUCLEAR REACTOR

    DOE Patents [OSTI]

    Teitel, R.J.

    1961-09-01

    A control means is described for a reactor which employs a liquid fuel consisting of a fissile isotope in a liquid bismuth solvent. The liquid fuel is contained in a plurality of tubular vessels. Control is effected by inserting plungers in the vessels to displace the liquid fuel and provide a critical or non- critical fuel configuration as desired.

  17. PDET-A New Tool for Partial Defect Verification of Pressurized...

    Office of Scientific and Technical Information (OSTI)

    of Pressurized Water Reactor Spent Fuel Assemblies Citation Details In-Document Search Title: PDET-A New Tool for Partial Defect Verification of Pressurized Water Reactor Spent ...

  18. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  19. Repair of hydroprocessing reactors

    SciTech Connect (OSTI)

    Niccolls, E.H.; Imgram, A.G.; Bagdasarian, A.J.

    1995-12-01

    In this paper the authors very briefly review the types of damage observed and repairs performed in hydroprocessing reactors. The authors will summarize the fundamental damage mechanisms, and where in the reactors they have needed to make repairs over the years. The era from the 1960`s through about 1990 will be briefly discussed. They describe in more detail repairs undertaken in the 1990`s. Finally, they will note important technical issues their industry faces regarding repairs, and the long term reliable operation of these vessels.

  20. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  1. Pressure testing of torispherical heads

    SciTech Connect (OSTI)

    Rana, M.D.; Kalnins, A.; Updike, D.P.

    1995-12-01

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

  2. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    SciTech Connect (OSTI)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  3. Radioactive material release from a containment vessel during a fire accident

    SciTech Connect (OSTI)

    Hensel, S.; Norkus, J.

    2015-02-26

    A methodology is presented to determine the source term for leaks and ruptures of pressurized vessels. The generic methodology is applied to a 9975 Primary Containment Vessel (PCV) which losses containment due to a hypothesized fire accident. The release due to a vessel rupture is approximately two orders of magnitude greater than the release due to a leak.

  4. The Effect of Accident Conditions on the Molten Core Material Relocation into the Lower Head of a PWR Vessel with Application to TMI-2

    SciTech Connect (OSTI)

    An Xuegao; Dhir, Vijay K.; Okrent, David

    2000-11-15

    The damage progression of the reactor core and the slumping mechanism of molten material to the lower head of the reactor vessel were examined through simulation of severe accident scenarios that lead to large-scale core damage. The calculations were carried out on a Three Mile Island Unit 2 configuration using the computer code SCDAP/RELAP5/MOD3.2.Different accident scenarios were simulated. The high-pressure injection and makeup flow rates were changed. The extreme case with no water being added during the accident was examined. Reflood by restart of coolant pump 2B was also studied. Finally, the size of the power-operated relief valve opening was also changed. The effects of these accident scenarios on the accident progression and the core damage process were studied.It is concluded that, according to code MOD3.2, the molten material slumped to the lower head of the reactor vessel when the junction of the top and side crusts failed after the molten pool had reached the periphery of the core. When the effective stress caused by pressure imbalance inside and outside of the crust became larger than the ultimate strength of the crust, the crust failed mechanically.

  5. Pressure sensor for sealed containers

    DOE Patents [OSTI]

    Hodges, Franklin R.

    2001-01-01

    A magnetic pressure sensor for sensing a pressure change inside a sealed container. The sensor includes a sealed deformable vessel having a first end attachable to an interior surface of the sealed container, and a second end. A magnet mounted to the vessel second end defining a distance away from the container surface provides an externally detectable magnetic field. A pressure change inside the sealed container causes deformation of the vessel changing the distance of the magnet away from the container surface, and thus the detectable intensity of the magnetic field.

  6. Economic advantages of Division 2 design for vessels per ASME Code Section VIII

    SciTech Connect (OSTI)

    Lengsfeld, M.; Holman, R.; Lengsfeld, P.F.

    1995-12-01

    ASME Boiler and Pressure Vessel Code Section 8, Division 2 has been available since 1968 for the design of pressure equipment. Industry has generally accepted this code for the design of high pressure vessels, high pressure being relative. Some consider high pressure above 3,000 PSIG, others look at high pressure above 1,000 or 1,500 PSIG. There are organizations who tie the use of Division 2 to thickness, meaning vessels in a thickness range above 3 to 4 inches as worthwhile to design to Division 2. In this paper the authors discuss the use of Division 2 strictly as an economic issue. Independent of thickness, if say a 3/4 in. thick vessel is lower in cost designed to Division 2 vs Division 1 why would one not build this vessel using Division 2 as the design basis?

  7. Cyclic corrosion crack resistance curves of certain vessel steels

    SciTech Connect (OSTI)

    Panasyuk, V.V.; Fedorova, V.A.; Pusyak, S.A.; Ratych, L.V.; Timofeev, L.V.; Zuezdin, Y.I.

    1985-11-01

    Results are presented of investigations of 15Kh2MFA and 15Kh2NMFA steels. In the first stage of the investigations, the cyclic corrosion crack resistance characteristics were determined with limiting values of the various factors: loading frequency, loading cycle stress ratio, temperature and length of service. An intense flow of ionizing radiation may markedly change the mechanical properties in 30-40 years; this acts on the reactor vessel. The experimental data for strength categories KP-45 and KP-90 of both vessel steels lies in a quite narrow band of spread, which provides a basis for representing it by a single generalized curve, presented here. The result of cyclic corrosion crack resistance tests of disk specimens of 15Kh2MFA and 15Kh2NMFA vessel steels in boric acid controlled reactor water solution in distilled water with the addition of KOH to pH 8 was established.

  8. Analysis of a high pressure ATWS (anticipated transient without scram) with very low make-up flow

    SciTech Connect (OSTI)

    Wagner, K.C.

    1988-10-01

    A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs.

  9. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, V.

    1984-06-13

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  10. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, Vladimir (Skokie, IL)

    1986-01-01

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

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

    DOE Patents [OSTI]

    Janikowski, Stuart K. (Idaho Falls, ID)

    2000-01-01

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

  12. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B.; Altman, David A.; Singleton, Norman R.

    2011-02-15

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

  13. Heat barrier for use in a nuclear reactor facility

    DOE Patents [OSTI]

    Keegan, Charles P.

    1988-01-01

    A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

  14. Control system for a small fission reactor

    DOE Patents [OSTI]

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

    1985-02-08

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

  15. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect (OSTI)

    Tsiklauri, G.; Newman, D.; Meriwether, G.; Korolev, V. [Pacific Northwest National Laboratory, P.O. Box 999 Richland, WA 99352 (United States)

    2002-07-01

    An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO{sub 2} with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 deg. C elevates the steam cycle to high thermal efficiency of 42%. (authors)

  16. Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

    SciTech Connect (OSTI)

    Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

    2015-03-15

    Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

  17. Current status of the advanced high temperature reactor

    SciTech Connect (OSTI)

    Holcomb, D. E.; Iias, D.; Quails, A. L.; Peretz, F. J.; Varma, V. K.; Bradley, E. C.; Cisneros, A. T.

    2012-07-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Dept. of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. (authors)

  18. Hanging core support system for a nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

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

  19. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

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

  20. Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

    SciTech Connect (OSTI)

    Glenn E McCreery; Keith G Condie

    2006-09-01

    The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

  1. Laser-driven fusion reactor

    DOE Patents [OSTI]

    Hedstrom, J.C.

    1973-10-01

    A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)

  2. OSS 19.4 Pressure Safety 3/27/95

    Broader source: Energy.gov [DOE]

    The objective of this surveillance is to evaluate the contractor's implementation of programs to ensure the integrity of pressure vessels and minimize risks from failure of vessels to the public...

  3. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect (OSTI)

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

    1993-01-01

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

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  5. High pressure furnace

    DOE Patents [OSTI]

    Morris, D.E.

    1993-09-14

    A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum)). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior. 19 figures.

  6. High pressure furnace

    DOE Patents [OSTI]

    Morris, Donald E.

    1993-01-01

    A high temperature high pressure furnace has a hybrid partially externally heated construction. A metallic vessel fabricated from an alloy having a composition of at least 45% nickel, 15% chrome, and 10% tungsten is utilized (the preferred alloy including 55% nickel, 22% chrome, 14% tungsten, 2% molybdenum, 3% iron (maximum) and 5% cobalt (maximum). The disclosed alloy is fabricated into 11/4 or 2 inch, 32 mm or 50 mm bar stock and has a length of about 22 inches, 56 cm. This bar stock has an aperture formed therein to define a closed high temperature, high pressure oxygen chamber. The opposite and closed end of the vessel is provided with a small blind aperture into which a thermocouple can be inserted. The closed end of the vessel is inserted into an oven, preferably heated by standard nickel chrome electrical elements and having a heavily insulated exterior.

  7. High pressure liquid level monitor

    DOE Patents [OSTI]

    Bean, Vern E.; Long, Frederick G.

    1984-01-01

    A liquid level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  8. Ultrasonic examination of the heavy section test blocks for hydroprocessing reactors used by the petroleum industry

    SciTech Connect (OSTI)

    Dodd, F.J.; Zhang, Y.; Imgram, A.

    1995-12-01

    An 18-inch (457 mm) thick Cr-Mo steel test block with an austenitic stainless steel ID cladding was designed and fabricated at the direction of the Material Properties Council in cooperation with the Pressure Vessel Research Committee (PVRC) Committee on Nondestructive Examination of Components. The test block contains several embedded flaws located near both ID and OD surfaces. Chevron Research and Technology Company and WesDyne International conducted the initial UT Examinations on this block at Chevron. In addition, a 10-inch (254 mm) thick block with embedded flaws was also examined. All the flaws m the two blocks were either cracks or lack of fusion. The examination results demonstrate the feasibility of UT to detect and size flaws in hydroprocessing reactor vessels and provide a basis to compare the relative capabilities of UT with the radiographic (RT) inspections currently required by the fabrication code.

  9. Radial pressure flange seal

    DOE Patents [OSTI]

    Batzer, Thomas H.; Call, Wayne R.

    1989-01-01

    This invention provides an all metal seal for vacuum or pressure vessels or systems. This invention does not use gaskets. The invention uses a flange which fits into a matching groove. Fluid pressure is applied in a chamber in the flange causing at least one of the flange walls to radially press against a side of the groove creating the seal between the flange wall and the groove side.

  10. Radial pressure flange seal

    DOE Patents [OSTI]

    Batzer, T.H.; Call, W.R.

    1989-01-24

    This invention provides an all metal seal for vacuum or pressure vessels or systems. This invention does not use gaskets. The invention uses a flange which fits into a matching groove. Fluid pressure is applied in a chamber in the flange causing at least one of the flange walls to radially press against a side of the groove creating the seal between the flange wall and the groove side. 5 figs.

  11. Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

    SciTech Connect (OSTI)

    Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin; Chu, In-Cheol; Song, Chul-Hwa [Korea Atomic Energy Research Institute (Korea, Republic of)

    2003-07-15

    Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focused on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.

  12. PRESSURE RELIEF DEVICE DATA SHEET FORM PS-5 Pressure System Number: Date:

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

    RELIEF DEVICE DATA SHEET FORM PS-5 Pressure System Number: Date: Pressure System Name: Pressure Vessel Number (if Applicable): Device installed directly on vessel?: __Yes __No Code: System Fluid: Code Year: Fluid State: Fluid Category: RELIEF DEVICE DATA Device Type ___Safety Relief Valve ____Rupture Disk ___Other (specify) Certification Type: ___ASME ___CE/PED ___Other (specify) Manufacturer Rated Flow Capacity: Part Number Converted Flow Capacity: Serial Number Set Pressure Inspection/Test

  13. Automatic safety rod for reactors. [LMFBR

    DOE Patents [OSTI]

    Germer, J.H.

    1982-03-23

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

  14. Vessel structural support system

    DOE Patents [OSTI]

    Jenko, James X.; Ott, Howard L.; Wilson, Robert M.; Wepfer, Robert M.

    1992-01-01

    Vessel structural support system for laterally and vertically supporting a vessel, such as a nuclear steam generator having an exterior bottom surface and a side surface thereon. The system includes a bracket connected to the bottom surface. A support column is pivotally connected to the bracket for vertically supporting the steam generator. The system also includes a base pad assembly connected pivotally to the support column for supporting the support column and the steam generator. The base pad assembly, which is capable of being brought to a level position by turning leveling nuts, is anchored to a floor. The system further includes a male key member attached to the side surface of the steam generator and a female stop member attached to an adjacent wall. The male key member and the female stop member coact to laterally support the steam generator. Moreover, the system includes a snubber assembly connected to the side surface of the steam generator and also attached to the adjacent wall for dampening lateral movement of the steam generator. In addition, the system includes a restraining member of "flat" attached to the side surface of the steam generator and a bumper attached to the adjacent wall. The flat and the bumper coact to further laterally support the steam generator.

  15. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  16. Integrated reformer and shift reactor

    DOE Patents [OSTI]

    Bentley, Jeffrey M.; Clawson, Lawrence G.; Mitchell, William L.; Dorson, Matthew H.

    2006-06-27

    A hydrocarbon fuel reformer for producing diatomic hydrogen gas is disclosed. The reformer includes a first reaction vessel, a shift reactor vessel annularly disposed about the first reaction vessel, including a first shift reactor zone, and a first helical tube disposed within the first shift reactor zone having an inlet end communicating with a water supply source. The water supply source is preferably adapted to supply liquid-phase water to the first helical tube at flow conditions sufficient to ensure discharge of liquid-phase and steam-phase water from an outlet end of the first helical tube. The reformer may further include a first catalyst bed disposed in the first shift reactor zone, having a low-temperature shift catalyst in contact with the first helical tube. The catalyst bed includes a plurality of coil sections disposed in coaxial relation to other coil sections and to the central longitudinal axis of the reformer, each coil section extending between the first and second ends, and each coil section being in direct fluid communication with at least one other coil section.

  17. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-03-02

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

  18. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01

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

  19. Single module pressurized fuel cell turbine generator system

    DOE Patents [OSTI]

    George, Raymond A.; Veyo, Stephen E.; Dederer, Jeffrey T.

    2001-01-01

    A pressurized fuel cell system (10), operates within a common pressure vessel (12) where the system contains fuel cells (22), a turbine (26) and a generator (98) where preferably, associated oxidant inlet valve (52), fuel inlet valve (56) and fuel cell exhaust valve (42) are outside the pressure vessel.

  20. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect (OSTI)

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  1. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  2. A method and apparatus for destroying hazardous organics and other combustible materials in a subcritical/supercritical reactor

    SciTech Connect (OSTI)

    Janikowski, Stuart K.

    1997-12-01

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

  3. Decommissioning of the BR3 reactor: status and perspectives

    SciTech Connect (OSTI)

    Noynaert, L.; Verstraeten, I.

    2007-07-01

    The BR3 plant at Mol in Belgium built at the end of the fifties was the first PWR plant built outside the USA. The reactor had a small net power output (10 MWe) but comprised all the loops and features of a commercial PWR plant. The BR3 plant was operated with the main objective of testing advanced PWR fuels under irradiation conditions similar to those encountered in large commercial PWR plants. The reactor was started in 1962 and shut down in 1987 after 25 years of continuous operation. Since 1989, SCK.CEN is decommissioning the BR3 PWR research reactor. The dismantling of the metallic components including reactor pressure vessel and internals is completed and extensively reported in the literature. The dismantling of auxiliary components and the decontamination of parts of the infrastructure are now going on. The decommissioning progress is continuously monitored and costs and strategy are regularly reassessed. The first part of the paper describes the main results and lessons learned from the reassessment exercises performed in 1994, 1999, 2004 and 2007. Impacts of changes in legal framework on the decommissioning costs will be addressed. These changes concern e.g. licensing aspects, clearance levels, waste management... The middle part of the paper discusses the management of activated and/or contaminated concrete. The costing exercise performed in 1995 highlighted that the management of activated and contaminated concrete is the second main cost item after the dismantling of the reactor pressure vessel and internals. Different possible solutions were studied. These are evacuation as radioactive waste with or without supercompaction, recycling this 'radioactive' grout or concrete for conditioning of radioactive waste e.g. conditioning of metallic waste. The paper will give the results of the cost-benefit analysis made to select the solution retained. The last part of the paper will discuss the end goal of the decommissioning of the BR3. In the final

  4. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-04-04

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

  5. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

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

  6. HOMOGENEOUS NUCLEAR REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  7. Hydrogasification reactor and method of operating same

    DOE Patents [OSTI]

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

    2013-09-10

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

  8. Design and optimization of a back-flow limiter for the high performance light water reactor

    SciTech Connect (OSTI)

    Fischer, Kai; Laurien, Eckart; Claas, Andreas G.; Schulenberg, Thomas

    2007-07-01

    Design and Analysis of a back-flow limiter are presented, which is implemented as a safety device in the four inlet lines of the Reactor Pressure Vessel (RPV) of the High Performance Light Water Reactor (HPLWR). As a passive component, the back-flow limiter has no moving parts and belongs to the group of fluid diodes. It has low flow resistance for regular operation condition and a high flow resistance when the flow direction is reversed which is the case if a break of the feedwater line occurs. The increased flow resistance is due to a substantially increased swirl for reverse flow condition. The design is optimized employing 1D flow analyses in combination with 3D CFD analyses with respect to geometrical modifications, like the nozzle shape and swirler angles. (authors)

  9. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOE Patents [OSTI]

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  10. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOE Patents [OSTI]

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  11. Bonfire Tests of High Pressure Hydrogen Storage Tanks

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  12. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  13. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect (OSTI)

    Martin, Scott; Rood, Marc

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  14. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

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

    1984-08-30

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

  15. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

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

  16. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  17. Ultrasonic liquid-level detector for varying temperature and pressure environments

    DOE Patents [OSTI]

    Anderson, R.L.; Miller, G.N.

    1981-10-26

    An ultrasonic liquid level detector for use in varying temperature and pressure environments, such as a pressurized water nuclear reactor vessel, is provided. The detector employs ultrasonic extensional and torsional waves launched in a multiplexed alternating sequence into a common sensor. The sensor is a rectangular cross section stainless steel rod which extends into the liquid medium whose level is to be detected. The sensor temperature derived from the extensional wave velocity measurements is used to compensate for the temperature dependence of the torsional wave velocity measurements which are also level dependent. The torsional wave velocity measurements of a multiple reflection sensor then provide a measurement of liquid level over a range of several meters with a small uncertainty over a temperature range of 20 to 250/sup 0/C and pressures up to 15 MPa.

  18. Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.

    1977-01-01

    A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the

  19. Method and apparatus for detecting irregularities on or in the wall of a vessel

    DOE Patents [OSTI]

    Bowling, Michael Keith (Blackborough Cullompton, GB)

    2000-09-12

    A method of detecting irregularities on or in the wall of a vessel by detecting localized spatial temperature differentials on the wall surface, comprising scanning the vessel surface with a thermal imaging camera and recording the position of the or each region for which the thermal image from the camera is indicative of such a temperature differential across the region. The spatial temperature differential may be formed by bacterial growth on the vessel surface; alternatively, it may be the result of defects in the vessel wall such as thin regions or pin holes or cracks. The detection of leaks through the vessel wall may be enhanced by applying a pressure differential or a temperature differential across the vessel wall; the testing for leaks may be performed with the vessel full or empty, and from the inside or the outside.

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

    SciTech Connect (OSTI)

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

    2004-03-01

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