HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

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

75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-09

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will... would result in major inconvenience. Dated: March 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch...

75 FR 71744 - Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0361] Toshiba Corporation Power Systems Company Notice of Receipt and Availability of an Application for Renewal of the U.S. Advanced Boiling Water Reactor Design... application for a design certification (DC) renewal for the U.S. Advanced Boiling Water Reactor (ABWR). An...

76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

The startup of the Dodewaard natural circulation boiling water reactor -- Experiences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nissen, W.H.M.; Van Der Voet, J.; Karuza, J.

1994-07-01

Because of its similarity to the simplified boiling water reactor (SBWR), the Dodewaard natural circulation boiling water reactor (BWR) is of special interest to further development of the SBWR design. It has become especially important to gain more insight into the Dodewaard BWR behavior during startup, paying special attention to its stability. Therefore, special instrumentation was used by means of which a series of measurements were taken during the two startups in February and June 1992. The results obtained from these measurements are used to deepen insight into the recirculation flow and the stability of the reactor during startup undermore » conditions with a normal pressure/power trajectory. They have already shown a very early recirculation flow onset during low-power operation and no indication of reactor instability. Furthermore, they will be used as a basis for the research program investigating the reactor behavior under different pressure/power conditions, which is scheduled for next year.« less

Preliminary design and hazards report. Boiling Reactor Experiment V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R. E.

1960-02-01

The preliminary objectives of the proposed BORAX V program are to test nuclear superheating concepts and to advance the technology of boiling-water-reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities. The reactor vessel is a cylinder with ellipsoidal heads, made of carbon steel clad internally with stainless steel. Each of the three cores is 24 in. high and has an effective diameter of 39 in. This is a preliminary report. (W.D.M.)

BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

DOEpatents

Treshow, M.

1963-04-30

This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

CHIMNEY FOR BOILING WATER REACTOR

DOEpatents

Petrick, M.

1961-08-01

A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-14

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0128] All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident...

BOILING SLURRY REACTOR AND METHOD FO CONTROL

DOEpatents

Petrick, M.; Marchaterre, J.F.

1963-05-01

The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.; Bentz, J.H.; Simpson, R.B.

1995-06-01

Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation.

Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.; Slezak, S.E.; Bentz, J.H.

1994-03-01

This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm{sup 2} across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactormore » vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests.« less

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wheeler, Timothy A.; Liao, Huafei

2014-12-01

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constitutedmore » a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.« less

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-11

...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

DOEpatents

Tokarz, R.D.

1981-10-27

This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

DOEpatents

Holl, R.J.; Klecker, R.W.; Graham, C.B.

1962-05-15

A description is given of a boiling water reactor having a superheating region integral with the core. The core consists essentially of an annular boiling region surrounding an inner superheating region. Both regions contain fuel elements and are separated by a cylindrical wall, perforations being provided in the lower portion of the cylindrical wall to permit circulation of a common water moderator between the two regions. The superheater region comprises a plurality of tubular fuel assemblies through which the steam emanating from the boiling region passes to the steam outlet. Each superheater fuel assembly has an outer double-walled cylinder, the double walls being concentrically spaced and connected together at their upper ends but open at the bottom to provide for differential thermal expansion of the inner and outer walls. Gas is entrapped in the annulus between the walls which acts as an insulating space between the fissionable material inside and the moderator outside. (AEC)

(Boiling water reactor (BWR) CORA experiments)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.

To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of themore » BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.« less

Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-03-01

This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 inmore » their evaluation process. 73 refs., 26 figs., 69 tabs.« less

75 FR 21046 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on May 6-8, 2010, 11545 Rockville Pike, Rockville....: Boiling Water Reactor (BWR) Owners Group (BWROG) Topical Report NEDC-33347P, ``Containment Overpressure...

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, D.M.

1996-03-12

An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

Development of Advanced ISS-WPA Catalysts for Organic Oxidation at Reduced Pressure/Temperature

NASA Technical Reports Server (NTRS)

Yu, Ping; Nalette, Tim; Kayatin, Matthew

2016-01-01

The Water Processor Assembly (WPA) at International Space Station (ISS) processes a waste stream via multi-filtration beds, where inorganic and non-volatile organic contaminants are removed, and a catalytic reactor, where low molecular weight organics not removed by the adsorption process are oxidized at elevated pressure in the presence of oxygen and elevated temperature above the normal water boiling point. Operation at an elevated pressure requires a more complex system design compared to a reactor that could operate at ambient pressure. However, catalysts currently available have insufficient activity to achieve complete oxidation of the organic load at a temperature less than the water boiling point and ambient pressure. Therefore, it is highly desirable to develop a more active and efficient catalyst at ambient pressure and a moderate temperature that is less than water boiling temperature. This paper describes our efforts in developing high efficiency water processing catalysts. Different catalyst support structures and coating metals were investigated in subscale reactors and results were compared against the flight WPA catalyst. Detailed improvements achieved on alternate metal catalysts at ambient pressure and 200 F will also be presented in the paper.

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

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

DOEpatents

Hill, Paul R.

1994-01-01

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

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

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)

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.

The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

77 FR 69900 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on December 6-8, 2012, 11545 Rockville... Recommendations (SECY-12-0064), (3) Venting Systems for Boiling Water Reactors (BWRs) with Mark I and Mark II...

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

75 FR 55365 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-10

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Joint Subcommittee The ACRS Subcommittees on Thermal Hydraulics Phenomena; Advanced Boiling Water Reactor (ABWR); and Materials, Metallurgy, and Reactor Fuels will hold a joint meeting on October 4, 2010...

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

DOEpatents

Hill, P.R.

1994-12-27

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

Nuclear reactor with low-level core coolant intake

DOEpatents

Challberg, Roy C.; Townsend, Harold E.

1993-01-01

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

77 FR 41814 - Entergy Operations, Inc.; Grand Gulf Nuclear Station, Unit 1

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-16

... Unit 1 result primarily from periodic testing of diesel generators and fire water pump diesel engines... rural. GGNS Unit 1 is a General Electric Mark 3 boiling-water reactor. Identification of the Proposed... following: replacing the reactor feed pump turbine rotors; replacing the main generator current transformers...

Nuclear Power from Fission Reactors. An Introduction.

ERIC Educational Resources Information Center

Department of Energy, Washington, DC. Technical Information Center.

The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... facility consists of a boiling-water reactor located in West Feliciana Parish, Louisiana. 2.0 Request... Containment Leakage Testing for Water- Cooled Power Reactors,'' requires that components which penetrate containment be periodically leak tested at the ``P a, '' defined as the ``calculated peak containment internal...

Equalization of energy density in boiling water reactors (as exemplified by WB-50). Development and testing of WB -50 computational model on the basis of MCU-RR code

NASA Astrophysics Data System (ADS)

Chertkov, Yu B.; Disyuk, V. V.; Pimenov, E. Yu; Aksenova, N. V.

2017-01-01

Within the framework of research in possibility and prospects of power density equalization in boiling water reactors (as exemplified by WB-50) a work was undertaken to improve prior computational model of the WB-50 reactor implemented in MCU-RR software. Analysis of prior works showed that critical state calculations have deviation of calculated reactivity exceeding ±0.3 % (ΔKef/Kef) for minimum concentrations of boric acid in the reactor water and reaching 2 % for maximum concentration values. Axial coefficient of nonuniform burnup distribution reaches high values in the WB-50 reactor. Thus, the computational model needed refinement to take into account burnup inhomogeneity along the fuel assembly height. At this stage, computational results with mean square deviation of less than 0.7 % (ΔKef/Kef) and dispersion of design values of ±1 % (ΔK/K) shall be deemed acceptable. Further lowering of these parameters apparently requires root cause analysis of such large values and paying more attention to experimental measurement techniques.

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paniagua, J.; Rohatgi, U.S.; Prasad, V.

1996-10-01

RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.

1998-04-01

For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

75 FR 16517 - Dominion Nuclear Connecticut, Inc.; Millstone Power Station, Unit Nos 1, 2, and 3; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

.... Borchardt, NRC, to M. S. Fertel, Nuclear Energy Institute, ADAMS Accession No. ML091410309). The licensee's... effect. The facility consists of one boiling water reactor and two pressurized water reactors located in... public. The supplemental January 12, 2010, letter contains, as an attachment, an environmental assessment...

Small Modular Reactors: The Army’s Secure Source of Energy?

DTIC Science & Technology

2012-03-21

significant advantages of SMRs is the minimal amount of carbon dioxide (greenhouse gases) that is released in conjunction with the lifecycle operations...moderator in these reactors as well as the cooling agent and the means by which heat is removed to produce steam for turning the turbines of the...separate water system to generate steam to turn a turbine which then produces electricity. In the second type of light water reactors, the boiling water

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

BOILING WATER REACTOR TECHNOLOGY STATUS OF THE ART REPORT. VOLUME II. WATER CHEMISTRY AND CORROSION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breden, C.R.

1963-02-01

Information concerning the corrosive effects of water in power reactor moderator-coolant systems is presented. The information is based on investigations reported in the unclassified literature believed to be fairly complete to 1959, but less complete since then. The material is presented in sections on water decomposition, water chemistry, materials corrosion, corrosion product deposits, and radioactivity. It is noted that the report is presented as a part of a continuing program in development of less expensive materials for use in reactors. (J.R.D.)

75 FR 17786 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Power Uprates...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-07

... Boiling Water Reactor Owners Group's (BWROG) topical report NEDC-33347P, ``Containment Overpressure Credit... for the Use of Containment Accident Pressure in Determining the NPSH Margin of ECCS and Containment...

75 FR 5357 - In the Matter of Entergy Nuclear Operations, Inc., et al.; Order Extending the Effectiveness of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-02

...). Pilgrim is a boiling water nuclear reactor that is owned by Entergy Nuclear and operated by ENO. The... Generating Unit No. 1 (IP1). IP1 is a pressurized water nuclear reactor that is owned by ENIP2 and maintained... nuclear reactors that are owned by ENIP2 and ENIP3, respectively, and operated by ENO. The facilities are...

Energy from the Atom. A Basic Teaching Unit on Energy. Revised.

ERIC Educational Resources Information Center

McDermott, Hugh, Ed.; Scharmann, Larry, Ed.

Recommended for grades 9-12 social studies and/or physical science classes, this 4-8 day unit focuses on four topics: (1) the background and history of atomic development; (2) two common types of nuclear reactors (boiling water and pressurized water reactors); (3) disposal of radioactive waste; and (4) the future of nuclear energy. Each topic…

77 FR 39521 - Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-03

... Methodology for Boiling Water Reactors, June 2011. To support use of Topical Report ANP-10307PA, Revision 0... the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will... Water Reactors with AREVA Topical Report ANP- 10307PA, Revision 0, ``AREVA MCPR Safety Limit Methodology...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

NASA Astrophysics Data System (ADS)

Semidotskiy, I. I.; Kurskiy, A. S.

2013-12-01

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

NASA Astrophysics Data System (ADS)

Butler, Thomas S.

Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

CONTROL SYSTEM

DOEpatents

Shannon, R.H.; Williamson, H.E.

1962-10-30

A boiling water type nuclear reactor power system having improved means of control is described. These means include provisions for either heating the coolant-moderator prior to entry into the reactor or shunting the coolantmoderator around the heating means in response to the demand from the heat engine. These provisions are in addition to means for withdrawing the control rods from the reactor. (AEC)

Noise source and reactor stability estimation in a boiling water reactor using a multivariate autoregressive model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kanemoto, S.; Andoh, Y.; Sandoz, S.A.

1984-10-01

A method for evaluating reactor stability in boiling water reactors has been developed. The method is based on multivariate autoregressive (M-AR) modeling of steady-state neutron and process noise signals. In this method, two kinds of power spectral densities (PSDs) for the measured neutron signal and the corresponding noise source signal are separately identified by the M-AR modeling. The closed- and open-loop stability parameters are evaluated from these PSDs. The method is applied to actual plant noise data that were measured together with artificial perturbation test data. Stability parameters identified from noise data are compared to those from perturbation test data,more » and it is shown that both results are in good agreement. In addition to these stability estimations, driving noise sources for the neutron signal are evaluated by the M-AR modeling. Contributions from void, core flow, and pressure noise sources are quantitatively evaluated, and the void noise source is shown to be the most dominant.« less

REFLECTOR CONTROL OF A BOILING-WATER REACTOR

DOEpatents

Treshow, M.

1962-05-22

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scott, R.L.; Gallaher, R.B.

1977-08-02

This bibliography contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1976. The report includes 1,253 abstracts that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. They are arranged alphabetically by reactor name and then chronologically for each reactor. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes ofmore » failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Three of the unique events that occurred during the year are reviewed in detail.« less

Annotated bibliography of safety-related occurrences in boiling-water nuclear power plants as reported in 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scott, R.L.; Gallaher, R.B.

1976-07-01

The bibliography presented contains 100-word abstracts of reports to the U.S. Nuclear Regulatory Commission concerning operational events that occurred at boiling-water reactor nuclear power plants in 1975. The report includes 1169 abstracts, arranged alphabetically by reactor name and then chronologically for each reactor, that describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. Key-word and permuted-title indexes are provided to facilitate location of the subjects of interest, and tables that summarize the information contained in the bibliography are provided. The information listed in the tables includes instrument failures, equipment failures, system failures, causes of failures,more » deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction). Seven of the unique events that occurred during the year are reviewed in detail.« less

75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-20

... Monday, October 14, 2009, (74 FR 52829-52830). Thursday, October 7, 2010, Conference Room T2-B1, Two....: Final Safety Evaluation Report Associated with the Economic Simplified Boiling Water Reactor (ESBWR..., Inc. regarding the final Safety Evaluation Report associated with the ESBWR design certification...

Taking a fresh look at boiling heat transfer on the road to improved nuclear economics and efficiency

DOE PAGES

Pointer, William David; Baglietto, Emilio

2016-05-01

Here, in the effort to reinvigorate innovation in the way we design, build, and operate the nuclear power generating stations of today and tomorrow, nothing can be taken for granted. Not even the seemingly familiar physics of boiling water. The Consortium for the Advanced Simulation of Light Water Reactors, or CASL, is focused on the deployment of advanced modeling and simulation capabilities to enable the nuclear industry to reduce uncertainties in the prediction of multi-physics phenomena and continue to improve the performance of today’s Light Water Reactors and their fuel. An important part of the CASL mission is the developmentmore » of a next generation thermal hydraulics simulation capability, integrating the history of engineering models based on experimental experience with the computing technology of the future.« less

77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

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

New PANDA Tests to Investigate Effects of Light Gases on Passive Safety Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Auban, O.; Candreia, P.

The large- scale thermal-hydraulic PANDA facility (located at PSI in Switzerland), has been used over the last few years for investigating different passive decay- heat removal systems and containment phenomena for the next generation of light water reactors (Simplified Boiling Water Reactor: SBWR; European Simplified Boiling Water Reactor: ESBWR; Siedewasserreaktor: SWR-1000). Currently, as part of the European Commission 5. EURATOM Framework Programme project 'Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling' (TEMPEST), a new series of tests is being planned in the PANDA facility to experimentally investigate the distribution of non-condensable gases inside the containment andmore » their effect on the performance of the 'Passive Containment Cooling System' (PCCS). Hydrogen release caused by the metal-water reaction in the case of a postulated severe accident will be simulated in PANDA by injecting helium into the reactor pressure vessel. In order to provide suitable data for Computational Fluid Dynamic (CFD) code assessment and improvement, the instrumentation in PANDA has been upgraded for the new tests. In the present paper, a detailed discussion is given of the new PANDA tests to be performed to investigate the effects of light gas on passive safety systems. The tests are scheduled for the first half of the year 2002. (authors)« less

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fan-Bill Cheung; Joy L. Rempe

2004-06-01

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

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

NASA Astrophysics Data System (ADS)

Yang, Jun

Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (<800 °C). With the upgrade and development of advanced power reactors, however, enhancing the nucleate boiling rate and its upper limit, Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface coatings by comparing the measurements with those for a plain vessel without coatings. An overall enhancement in nucleate boiling rates and CHF limits up to 100% were observed. Moreover, combination of data from quenching experiments and steady-state experiments produced new sets of boiling curves, which covered both the nucleate and transient boiling regimes with much greater accuracy. Beside the experimental work, a theoretical CHF model has also been developed by considering the vapor dynamics and the boiling-induced two-phase motions in three separate regions adjacent to the heating surface. The CHF model is capable of predicting the performance of micro-porous coatings with given particle diameter, porosity, media permeability and thickness. It is found that the present CHF model agrees favorably with the experimental data. Effects of an enhanced vessel/insulation structure on the local nucleate boiling rate and CHF limit have also been investigated experimentally. It is observed that the local two-phase flow quantities such as the local void fraction, quality, mean vapor velocity, mean liquid velocity, and mean vapor and liquid mass flow rates could have great impact on the local surface heat flux as boiling of water takes place on the vessel surface. An upward co-current two-phase flow model has been developed to predict the local two-phase flow behavior for different flow channel geometries, which are set by the design of insulation structures. It is found from the two-phase flow visualization experiments and the two-phase flow model calculations that the enhanced vessel/insulation structure greatly improved the steam venting process at the minimum gap location compared to the performance of thermal insulation structures without enhancement. Moveover, depending on the angular location, steady-state boiling experiments with the enhanced insulation design showed an enhancement of 1.8 to 3.0 times in the local critical heat flux. Finally, nucleate boiling and CHF correlations were developed based on the data obtained from various quenching and steady-state boiling experiments. Additionally, CHF enhancement factors were determined and examined to show the separate and integral effects of the two ERVC enhancement methods. When both vessel coating and insulation structure were used simultaneously, the integral effect on CHF enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods.

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baversten, B.; Linden, M.J.

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclearmore » overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.« less

Preliminary design study of small long life boiling water reactor (BWR) with tight lattice thorium nitride fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su'ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id

2014-09-30

Neutronic performance of small long-life boiling water reactors (BWR) with thorium nitride based fuel has been performed. A recent study conducted on BWR in tight lattice environments (with a lower moderator percentage) produces small power reactor which has some specifications, i.e. 10 years operation time, power density of 19.1 watt/cc and maximum excess reactivity of about 4%. This excess reactivity value is smaller than standard reactivity of conventional BWR. The use of hexagonal geometry on the fuel cell of BWR provides a substantial effect on the criticality of the reactor to obtain a longer operating time. Supported by a tightmore » concept lattice where the volume fraction of the fuel is greater than the moderator and fuel, Thorium Nitride give good results for fuel cell design on small long life BWR. The excess reactivity of the reactor can be reduced with the addition of gadolinium as burnable poisons. Therefore the hexagonal tight lattice fuel cell design of small long life BWR that has a criticality more than 20 years of operating time has been obtained.« less

Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCormack, K.E.; Gallaher, R.B.

1982-03-01

This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in themore » tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).« less

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Process for separating dissolved solids from a liquid using an anti-solvent and multiple effect evaporators

DOEpatents

Daniels, Edward J.; Jody, Bassam J.; Bonsignore, Patrick V.

1994-01-01

A process and system for treating aluminum salt cake containing water soluble halide salts by contacting the salt cake with water to dissolve water soluble halide salts forming a saturated brine solution. Transporting a portion of about 25% of the saturated brine solution to a reactor and introducing into the saturated brine solution at least an equal volume of a water-miscible low-boiling organic material such as acetone to precipitate a portion of the dissolved halide salts forming a three-phase mixture of an aqueous-organic-salt solution phase and a precipitated salt phase and an organic rich phase. The precipitated salt phase is separated from the other phases and the organic rich phase is recycled to the reactor. The remainder of the saturated brine solution is sent to a multiple effect evaporator having a plurality of stages with the last stage thereof producing low grade steam which is used to boil off the organic portion of the solution which is recycled.

Process for separating dissolved solids from a liquid using an anti-solvent and multiple effect evaporators

DOEpatents

Daniels, E.J.; Jody, B.J.; Bonsignore, P.V.

1994-07-19

A process and system are disclosed for treating aluminum salt cake containing water soluble halide salts by contacting the salt cake with water to dissolve water soluble halide salts forming a saturated brine solution. Transporting a portion of about 25% of the saturated brine solution to a reactor and introducing into the saturated brine solution at least an equal volume of a water-miscible low-boiling organic material such as acetone to precipitate a portion of the dissolved halide salts forming a three-phase mixture of an aqueous-organic-salt solution phase and a precipitated salt phase and an organic rich phase. The precipitated salt phase is separated from the other phases and the organic rich phase is recycled to the reactor. The remainder of the saturated brine solution is sent to a multiple effect evaporator having a plurality of stages with the last stage thereof producing low grade steam which is used to boil off the organic portion of the solution which is recycled. 3 figs.

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Maolong; Ryals, Matthew; Ali, Amir

2016-08-01

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentallymore » investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.« less

Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel

DOE PAGES

Forrest, Eric C.; Don, Sarah M.; Hu, Lin -Wen; ...

2016-02-29

The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750more » to 3000 kg/m 2s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.« less

Feasibility of Rectangular Concrete Pressure Vessels for Human Occupancy

DTIC Science & Technology

1990-07-01

incorporated: into the s 1.71 ! -1 rule-. 20 DISTRIBUTION/ AVAILABILIT ’- ,"r ABSTRACT 2’ ABSTRACT SECURITY CLASSIFICATION 7 JUNCLASSIFIED/UNLIMITED 0 SAME AS RPT...carried the end loads. Gas cooled reactors were never very popular in the US. Domestic utilities preferred boiling water reactors that operated at...Point Tower); 1975 - 9,000 psi in Chicago ( Water Tower Place); 1984 - 10,000 psi in Seattle (Century Square Bldg.); 1987 - 10,000 psi in Toronto

8. Innovative Technologies: Two-Phase Heat Transfer in Water-Based Nanofluids for Nuclear Applications Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buongiorno, Jacopo; Hu, Lin-wen

2009-07-31

Nanofluids are colloidal dispersions of nanoparticles in water. Many studies have reported very significant enhancement (up to 200%) of the Critical Heat Flux (CHF) in pool boiling of nanofluids (You et al. 2003, Vassallo et al. 2004, Bang and Chang 2005, Kim et al. 2006, Kim et al. 2007). These observations have generated considerable interest in nanofluids as potential coolants for more compact and efficient thermal management systems. Potential Light Water Reactor applications include the primary coolant, safety systems and severe accident management strategies, as reported in other papers (Buongiorno et al. 2008 and 2009). However, the situation of interestmore » in reactor applications is often flow boiling, for which no nanofluid data have been reported so far. In this project we investigated the potential of nanofluids to enhance CHF in flow boiling. Subcooled flow boiling heat transfer and CHF experiments were performed with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (≤ 0.1 % by volume) at atmospheric pressure. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient (HTC) are similar (within ±20%). The HTC increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. The CHF tests were conducted at 0.1 MPa and at three different mass fluxes (1500, 2000, 2500 kg/m2s) under subcooled conditions. The maximum CHF enhancement was 53%, 53% and 38% for alumina, zinc oxide and diamond, respectively, always obtained at the highest mass flux. A post-mortem analysis of the boiling surface reveals that its morphology is altered by deposition of the particles during nanofluids boiling. A confocal-microscopy-based examination of the test section revealed that nanoparticles deposition not only changes the number of micro-cavities on the surface, but also the surface wettability. A simple model was used to estimate the ensuing nucleation site density changes, but no definitive correlation between the nucleation site density and the heat transfer coefficient data could be found. Wettability of the surface was substantially increased for heater coupons boiled in alumina and zinc oxide nanofluids, and such wettability increase seems to correlate reasonably well with the observed marked CHF enhancement for the respective nanofluids. Interpretation of the experimental data was conducted in light of the governing surface parameters (surface area, contact angle, roughness, thermal conductivity) and existing models. It was found that no single parameter could explain the observed HTC or CHF phenomena.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zabelin, A.I.; Shmelev, V.E.

Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterizedmore » by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.« less

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1991-05-21

This patent describes an improved fuel assembly for a boiling water reactor. It comprises: a fuel channel; a lower tie plate; an upper tie plate; the lower tie plate and the upper tie plate defining a two-dimensional matrix; at least one water rod the fuel rods being partial length rods.

Thermal energy storage material thermophysical property measurement and heat transfer impact

NASA Technical Reports Server (NTRS)

Tye, R. P.; Bourne, J. G.; Destarlais, A. O.

1976-01-01

The thermophysical properties of salts having potential for thermal energy storage to provide peaking energy in conventional electric utility power plants were investigated. The power plants studied were the pressurized water reactor, boiling water reactor, supercritical steam reactor, and high temperature gas reactor. The salts considered were LiNO3, 63LiOH/37 LiCl eutectic, LiOH, and Na2B4O7. The thermal conductivity, specific heat (including latent heat of fusion), and density of each salt were measured for a temperature range of at least + or - 100 K of the measured melting point. Measurements were made with both reagent and commercial grades of each salt.

Disparities in nuclear power plant performance in the United States and the Federal Republic of Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hansen, K.F.; Winje, D.K.

This report presents data comparing the performance of light water reactors in the United States and the Federal Republic of Germany (FRG). The comparisons are made for the years 1980-1983 and include 21 Westinghouse Pressurized Water Reactors (PWRs), 22 General Electric Boiling Water Reactors (BWRs) in the US; and 6 Kraftwerk Union (KWU) PWRs and 4 KWJ BWRs in the FRG. Data on capacity losses are presented in a disaggregated form for scheduled outages, forced outages, and regulatory imposed outages. Further, within the scheduled and forced outages, the data is subdivided into losses associated with the nuclear island, the balancemore » of plant, or other causes.« less

Boiling-Water Reactor internals aging degradation study. Phase 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

New Departure from Nucleate Boiling model relying on first principle energy balance at the boiling surface

NASA Astrophysics Data System (ADS)

Demarly, Etienne; Baglietto, Emilio

2017-11-01

Predictions of Departure from Nucleate Boiling have been a longstanding challenge when designing heat exchangers such as boilers or nuclear reactors. Many mechanistic models have been postulated over more than 50 years in order to explain this phenomenon but none is able to predict accurately the conditions which trigger the sudden change of heat transfer mode. This work aims at demonstrating the pertinence of a new approach for detecting DNB by leveraging recent experimental insights. The new model proposed departs from all the previous models by making the DNB inception come from an energy balance instability at the heating surface rather than a hydrodynamic instability of the bubbly layer above the surface (Zuber, 1959). The main idea is to modulate the amount of heat flux being exchanged via the nucleate boiling mechanism by the wetted area fraction on the surface, thus allowing a completely automatic trigger of DNB that doesn't require any parameter prescription. This approach is implemented as a surrogate model in MATLAB in order to validate the principles of the model in a simple and controlled geometry. Good agreement is found with the experimental data leveraged from the MIT Flow Boiling at various flow regimes. Consortium for Advanced Simulation of Light Water Reactors (CASL).

System Study: Reactor Core Isolation Cooling 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-08

... status of LACBWR means that there are no longer interconnected operating systems which require security... System (ADAMS), which provides text and image files of NRC's public documents. If you do not have access...

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Chemical Characterization of Simulated Boiling Water Reactor Coolant

DTIC Science & Technology

1990-05-01

33 Table 3. 1: BCCL Sample Block Design Calculations ........................................... 45 Table 5.1: Gas Absorption...cover gas . The cool, degassed pure water is pumped through a regenerative heat exchanger and then through an electric feedwater heater. The feedwater is...POINTS DWCMRHEAT DOWNOMER---EXCHANGER CHEMICAL GAHP INJECTIOIN PUMP SYSTEM COIVER GAS IN-CLIRE SECTION CAGN TANK RECOMBINER! ______ DEMINERALIZER (Cic

Spectral measurements of direct and scattered gamma radiation at a boiling-water reactor site

NASA Astrophysics Data System (ADS)

Block, R. C.; Preiss, I. L.; Ryan, R. M.; Vargo, G. J.

1990-12-01

Quantitative surveys of direct and scattered gamma radiation emitted from the steam-power conversion systems of a boiling-water reactor and other on-site radiation sources were made using a directionally shielded HPGe gamma spectrometry system. The purpose of this study was to obtain data on the relative contributions and energy distributions of direct and scattered gamma radiation in the site environs. The principal radionuclide of concern in this study is 16N produced by the 16O(n,p) 16N reaction in the reactor coolant. Due to changes in facility operation resulting from the implementation of hydrogen water chemistry (HWC), the amount of 16N transported from the reactor to the main steam system under full power operation is excepted to increase by a factor of 1.2 to 5.0. This increase in the 16N source term in the nuclear steam must be considered in the design of new facilities to be constructed on site as well as the evaluation of existing facilities with repect to ALARA (As Low As Reasonably Achievable) dose limits in unrestricted areas. This study consisted of base-line measurements taken under normal BWR chemistry conditions in October, 1987 and a corresponding set taken under HWC conditions in July, 1988. Ground-level and elevated measurements, corresponding to second-story building height, were obtained. The primary conclusion of this study is that direct radiation from the steam-power conversion system is the predominant source of radiation in the site environs of this reactor and that air scattering (i.e. skyshine) does not appear to be significant.

BOILING REACTORS

DOEpatents

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)

A two-step method for developing a control rod program for boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1992-01-01

This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in amore » computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE PAGES

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa; ...

2016-09-07

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Cameron S.; Zhang, Hongbin; Kucukboyaci, Vefa

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was used to simulate a typical pressurized water reactor (PWR) full core response with 17x17 fuel assemblies for a main steam line break (MSLB) accident scenario with the most reactive rod cluster control assembly stuck out of the core. The accident scenario was initiated at the hot zero power (HZP) at the end of the first fuel cycle with return to power state points that were determined by amore » system analysis code and the most limiting state point was chosen for core analysis. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this way, 59 full core simulations were performed to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. The results show that this typical PWR core remains within MDNBR safety limits for the MSLB accident.« less

Optimization of a Boiling Water Reactor Loading Pattern Using an Improved Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2003-08-15

A search method based on genetic algorithms (GA) using deterministic operators has been developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). The search method uses an Improved GA operator, that is, crossover, mutation, and selection. The handling of the encoding technique and constraint conditions is designed so that the GA reflects the peculiar characteristics of the BWR. In addition, some strategies such as elitism and self-reproduction are effectively used to improve the search speed. LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and three-dimensional-dependent constraints have alwaysmore » necessitated the use of three-dimensional core simulators for BWRs, so that an optimization method is required for computational efficiency. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant applying the Haling technique. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-08-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

Main steam-line break core shroud loading calculations for BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1995-12-31

In July 1994, the U.S. Nuclear regulatory Commission sent out Generic Letter 94-03 to all boiling water reactors in the United States, informing them of intergranular stress corrosion cracking of core shrouds found in 2 reactors. The letter directed all to perform safety analysis of the BWR units. Penn State performed scoping calculations to determine the forces experienced by the core shroud during a main-stream line break transient.

Assessment of correlations and models for the prediction of CHF in water subcooled flow boiling

NASA Astrophysics Data System (ADS)

Celata, G. P.; Cumo, M.; Mariani, A.

1994-01-01

The present paper provides an analysis of available correlations and models for the prediction of Critical Heat Flux (CHF) in subcooled flow boiling in the range of interest of fusion reactors thermal-hydraulic conditions, i.e. high inlet liquid subcooling and velocity and small channel diameter and length. The aim of the study was to establish the limits of validity of present predictive tools (most of them were proposed with reference to light water reactors (LWR) thermal-hydraulic studies) in the above conditions. The reference dataset represents almost all available data (1865 data points) covering wide ranges of operating conditions in the frame of present interest (0.1 less than p less than 8.4 MPa; 0.3 less than D less than 25.4 mm; 0.1 less than L less than 0.61 m; 2 less than G less than 90.0 Mg/sq m/s; 90 less than delta T(sub sub,in) less than 230 K). Among the tens of predictive tools available in literature four correlations (Levy, Westinghouse, modified-Tong and Tong-75) and three models (Weisman and Ileslamlou, Lee and Mudawar and Katto) were selected. The modified-Tong correlation and the Katto model seem to be reliable predictive tools for the calculation of the CHF in subcooled flow boiling.

RELAP-7 Development Updates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J. R.; Bergeron, A.; Dionne, B.

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cmmore » 2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).« less

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Flowsheets and source terms for radioactive waste projections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

NASA Technical Reports Server (NTRS)

Sorensen, Kirk; Juhasz, Albert

2007-01-01

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-27

... used.'' In 10 CFR 73.55, entitled ``Requirements for physical protection of licensed activities in...-shutdown decommissioning activities report (PSDAR). The DPC has been conducting dismantlement and decommissioning activities. The DPC is developing an onsite independent spent fuel storage installation (ISFSI...

77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-27

... intruders and ensure the integrity of physical barriers or other components and functions of the onsite... decommissioning plan is considered in the post-shutdown decommissioning activities report (PSDAR). The DPC has been conducting dismantlement and decommissioning activities. The DPC is developing an onsite...

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core... acceptable to implement with regard to initial testing features of emergency core cooling systems (ECCSs) for...

76 FR 32238 - Detroit Edison Company, Fermi 2; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-03

..., to M. S. Fertel, Nuclear Energy Institute (ADAMS Accession No. ML091410309)). The licensee's request... Programmatic Environmental Assessment and Finding of No Significant Impact (76 FR 187) for the treatment of..., the Commission) now or hereafter in effect. The facility consists of one boiling-water reactor located...

75 FR 14208 - Entergy Nuclear Operations, Inc.; Pilgrim Nuclear Power Station; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-24

... for all operating nuclear power plants, but noted that the Commission's regulations provide mechanisms...: June 4, 2009, letter from R. W. Borchardt, NRC, to M. S. Fertel, Nuclear Energy Institute). The... hereafter in effect. The facility consists of a boiling-water reactor located in Plymouth County...

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset willmore » be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.« less

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Application of the DG-1199 methodology to the ESBWR and ABWR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinich, Donald A.; Gauntt, Randall O.; Walton, Fotini

2010-09-01

Appendix A-5 of Draft Regulatory Guide DG-1199 'Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors' provides guidance - applicable to RADTRAD MSIV leakage models - for scaling containment aerosol concentration to the expected steam dome concentration in order to preserve the simplified use of the Accident Source Term (AST) in assessing containment performance under assumed design basis accident (DBA) conditions. In this study Economic and Safe Boiling Water Reactor (ESBWR) and Advanced Boiling Water Reactor (ABWR) RADTRAD models are developed using the DG-1199, Appendix A-5 guidance. The models were run using RADTRAD v3.03. Low Populationmore » Zone (LPZ), control room (CR), and worst-case 2-hr Exclusion Area Boundary (EAB) doses were calculated and compared to the relevant accident dose criteria in 10 CFR 50.67. For the ESBWR, the dose results were all lower than the MSIV leakage doses calculated by General Electric/Hitachi (GEH) in their licensing technical report. There are no comparable ABWR MSIV leakage doses, however, it should be noted that the ABWR doses are lower than the ESBWR doses. In addition, sensitivity cases were evaluated to ascertain the influence/importance of key input parameters/features of the models.« less

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

76 FR 9614 - Notice of Availability of the Proposed Models for Plant-Specific Adoption of Technical...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-18

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0039] Notice of Availability of the Proposed Models for..., ``Technical Specifications End States, NEDC-32988-A,'' for Boiling Water Reactor Plants Using the Consolidated... Availability. SUMMARY: The NRC is announcing the availability of the model application (with model no...

78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-31

... impacts. The proposed action will involve no construction or major renovation of any buildings or... to potential non-radiological impacts, there will be no construction or renovation of buildings or... adequate protection of public health and safety and common defense and security.'' The licensee claims that...

76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-20

.... Plain Language IX. Voluntary Consensus Standards X. Finding of No Significant Environmental Impact... information (or its equivalent) which was originally developed by GE Nuclear Energy (GE) and approved by the... from GE Hitachi Nuclear Energy (GEH) to use the GE- developed U.S. ABWR proprietary information for...

76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-16

... Environmental Impact: Availability IX. Paperwork Reduction Act Statement X. Regulatory Analysis XI. Regulatory.... ABWR; one commenter, GE Hitachi Nuclear Energy (GEH), was against the proposed amendment to the U.S... information from both the DCD developed by GE Nuclear Energy (GE) and the DCD developed by the STPNOC. The...

75 FR 15749 - Entergy Operations, Inc., Grand Gulf Nuclear Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-30

... request to extend the rule's compliance date for all operating nuclear power plants, but noted that the... (Nuclear Energy Institute) dated June 4, 2009. The licensee's request for an exemption is therefore...) now or hereafter in effect. The facility consists of a boiling-water reactor located in Claiborne...

75 FR 14209 - Entergy Nuclear Operations, Inc.; Vermont Yankee Nuclear Power Station; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-24

... compliance date for all operating nuclear power plants, but noted that the Commission's regulations provide...: June 4, 2009, letter from R.W. Borchardt, NRC, to M.S. Fertel, Nuclear Energy Institute). The licensee... Commission) now or hereafter in effect. The facility consists of a boiling-water reactor located in Windham...

High Fidelity BWR Fuel Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Su Jong

This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fractionmore » and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.« less

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

ORNL rod-bundle heat-transfer test data. Volume 2. Thermal-Hydraulic Test Facility experimental data report for test 3. 03. 6AR - transient film boiling in upflow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mullins, C. B.; Felde, D. K.; Sutton, A. G.

1982-04-01

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) Test 3.03.6AR. This test was conducted by members of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on May 21, 1980. Objective was to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small and large break loss-of-coolant accidents. Test 3.03.6AR was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.03.6AR available. Included in the report are uncertainties in the instrument responses,more » calculated mass flows, and calculated rod powers.« less

PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohr, C.L.; Pankaskie, P.J.; Heasler, P.G.

Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFITmore » model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

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

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

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

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

Expert system for maintenance management of a boiling water reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hong Shen; Liou, L.W.; Levine, S.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, whichmore » must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.« less

TRACE/PARCS Analysis of ATWS with Instability for a MELLLA+BWR/5

DOE PAGES

L. Y. Cheng; Baek, J. S.; Cuadra, A.; ...

2016-06-06

A TRACE/PARCS model has been developed to analyze anticipated transient without SCRAM (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands allowable operation in the power/flow map of a BWR to low flow rates at high power conditions. Such operation exacerbates the likelihood of large amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural circulation flow in the reactor pressure vessel (RPV) after the trip of the recirculationmore » pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heat-up that could result in localized fuel damage. TRACE predicts the heat-up to occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet; and the fuel becomes “locked” into a film boiling regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.« less

Real time monitoring of environmental crack growth in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, D.; Diehl, C.G.

1988-01-01

A comprehensive field test program was recently completed at several Boiling Water Reactors (BWR) to quantify the effect of coolant impurities on the initiation and growth of stress corrosion cracks. A new technology was utilized which allows for real time monitoring of stress corrosion crack growth rates. The BWR environments were characterized using Ion Chromatography and Electro Chemical Potential (ECP) measurements. The effects of typical water chemistry transients and startups were quantified.

Reactor for exothermic reactions

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

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.

Reactor for exothermic reactions

DOEpatents

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.

77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-19

... height, which inundated the Fukushima Dai-ichi nuclear power plant site. The earthquake and tsunami... and industry in the northeastern coastal areas of Japan. When the earthquake occurred, Fukushima Dai... Nuclear Accident at the Fukushima Daiichi Nuclear Power Station, Revision 0,'' issued November 2011, p. 72...

75 FR 9620 - Southern Nuclear Operating Company, Inc.; Edwin I. Hatch Nuclear Plant, Units 1 and 2; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-03

... industry request to extend the rule's compliance date for all operating nuclear power plants, but noted... M.S. Fertel, Nuclear Energy Institute). The licensee's request for an exemption is therefore... effect. The facility consists of two boiling-water reactors located in Appling County, Georgia. 2.0...

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

1995-09-01

The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

Optimization of Boiling Water Reactor Loading Pattern Using Two-Stage Genetic Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, Yoko; Aiyoshi, Eitaro

2002-10-15

A new two-stage optimization method based on genetic algorithms (GAs) using an if-then heuristic rule was developed to generate optimized boiling water reactor (BWR) loading patterns (LPs). In the first stage, the LP is optimized using an improved GA operator. In the second stage, an exposure-dependent control rod pattern (CRP) is sought using GA with an if-then heuristic rule. The procedure of the improved GA is based on deterministic operators that consist of crossover, mutation, and selection. The handling of the encoding technique and constraint conditions by that GA reflects the peculiar characteristics of the BWR. In addition, strategies suchmore » as elitism and self-reproduction are effectively used in order to improve the search speed. The LP evaluations were performed with a three-dimensional diffusion code that coupled neutronic and thermal-hydraulic models. Strong axial heterogeneities and constraints dependent on three dimensions have always necessitated the use of three-dimensional core simulators for BWRs, so that optimization of computational efficiency is required. The proposed algorithm is demonstrated by successfully generating LPs for an actual BWR plant in two phases. One phase is only LP optimization applying the Haling technique. The other phase is an LP optimization that considers the CRP during reactor operation. In test calculations, candidates that shuffled fresh and burned fuel assemblies within a reasonable computation time were obtained.« less

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

Silicon carbide composite for light water reactor fuel assembly applications

NASA Astrophysics Data System (ADS)

Yueh, Ken; Terrani, Kurt A.

2014-05-01

The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.

1994-04-01

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

75 FR 80854 - Toshiba Corporation; Acceptance for Docketing of an Application for Renewal of the U.S. Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-23

... NUCLEAR REGULATORY COMMISSION [Docket No. 52-044; NRC-2010-0361] Toshiba Corporation; Acceptance for Docketing of an Application for Renewal of the U.S. Advanced Boiling Water Reactor Design Certification On November 2, 2010, Toshiba Corporation (Toshiba) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for a design...

BWR Steam Dryer Alternating Stress Assessment Procedures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morante, R. J.; Hambric, S. A.; Ziada, S.

2016-12-01

This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2012 CFR

2012-01-01

...) of 10 CFR 50.34—Post-Accident Sampling for Boron, Chloride, and Dissolved Gases; and 3. Paragraph (f... affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if: (1) There is a substantial increase in the probability of an ex-vessel...

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

Etherification process

DOEpatents

Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

1990-01-01

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.

Oligomerization process

DOEpatents

Smith, Jr., Lawrence A.; hearn, Dennis; Jones, Jr., Edward M.

1991-01-01

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 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.

Best-estimate coupled RELAP/CONTAIN analysis of inadvertent BWR ADS valve opening transient

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Muftuoglu, A.K.

1993-01-01

Noncondensible gases may become dissolved in boiling water reactor (BWR) water-level instrumentation during normal operations. Any dissolved noncondensible gases inside these water columns may come out of solution during rapid depressurization events and displace water from the reference leg piping, resulting in a false high level. Significant errors in water-level indication are not expected to occur until the reactor pressure vessel (RPV) pressure has dropped below [approximately]450 psig. These water level errors may cause a delay or failure in emergency core cooling system (ECCS) actuation. The RPV water level is monitored using the pressure of a water column having amore » varying height (reactor water level) that is compared to the pressure of a water column maintained at a constant height (reference level). The reference legs have small-diameter pipes with varying lengths that provide a constant head of water and are located outside the drywell. The amount of noncondensible gases dissolved in each reference leg is very dependent on the amount of leakage from the reference leg and its geometry and interaction of the reactor coolant system with the containment, i.e., torus or suppression pool, and reactor building. If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response. In the postulated inadvertent opening of all seven automatic depressurization system (ADS) valves, the ECCS signal on high drywell pressure would be circumvented because the ADS valves discharge directly into the suppression pool. A best-estimate analysis of such an inadvertent opening of all ADS valves would have to consider the thermal-hydraulic coupling between the pool, drywell, reactor building, and RPV.« less

Divers muscle Fitzpatrick`s mussels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hobbs, B.; Kahabka, J.

1996-01-01

This article describes how an effective manual cleaning technique has helped rid submerged intake structures of this passive-aggressive pest. The New York Power Authority`s (NYPA) James A. Fitzpatrick (JAF) Nuclear Power Plant is located in Lycoming, NY, on the southeast shore of Lake Ontario. An 850-MWe, GE-design boiling water reactor (BWR), the unit has been in service since 1975. Water drawn from the Lake supplies cooling to plant loads via the circulating water system and three service water systems. These share a common intake system consisting of an offshore cap (crib), horizontal intake tunnel, two vertical risers and forebay/screenwell area.

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, Willem Jan; Marquino, Wayne

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

Dual-phase reactor plant with partitioned isolation condenser

DOEpatents

Hui, Marvin M.

1992-01-01

A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2013 CFR

2013-01-01

....34—Separate Plant Safety Parameter Display Console; 2. Paragraph (f)(2)(viii) of 10 CFR 50.34—Post... design bases or in the safety analyses. c. A proposed departure from Tier 2 affecting resolution of an ex... if: (1) There is a substantial increase in the probability of an ex-vessel severe accident such that...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2014 CFR

2014-01-01

....34—Separate Plant Safety Parameter Display Console; 2. Paragraph (f)(2)(viii) of 10 CFR 50.34—Post... design bases or in the safety analyses. c. A proposed departure from Tier 2 affecting resolution of an ex... if: (1) There is a substantial increase in the probability of an ex-vessel severe accident such that...

Etherification process

DOEpatents

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

1990-08-21

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. 2 figs.

Method for conducting exothermic reactions

DOEpatents

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

1993-01-05

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 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.

Method for conducting exothermic reactions

DOEpatents

Smith, Jr., Lawrence; Hearn, Dennis; Jones, Jr., Edward M.

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.

Oligomerization process

DOEpatents

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

1991-03-26

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. 2 figures.

The behavior of breached boiling water reactor fuel rods on long-term exposure to air and argon at 598 K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohli, R.; Gilbert, E.R.; Johnson, A.B.

1985-05-01

Two irradiated boiling water reactor fuel rods with breached cladding were exposed to argon and to air at 598 K for 7.56 Ms (2100 h). These tests were conducted to determine fuel swelling and cladding crack propagation under conditions that promote UO/sub 2/ fuel oxidation and to observe the behavior of water-logged breached fuel in an inert gas environment. The two rods were selected for testing after extensive hot cell examination had shown the cladding of both rods to be breached with several centimetres of open cracks; the cracks were characterized in detail before the test. As part of themore » experiment, the amount of the readily removable water contained in the fuel rods was determined. To oxidize the fuel to a significant level ( about10%), the air in the annealine capsule was replenished approximately daily. The depletion of oxygen available in the air capsule due to fuel oxidation occurred in about0.036 Ms (10 h). At the end of the test period, about6% of the fuel is estimated to have oxidized. Posttest examination of the rods showed that cladding degradation resulted from swelling due to oxidation of the fuel in the air environment. The cladding degradation was localized and fuel oxidation did not measurably extend beyond the cladding breach. No cladding degradation was measurable in the breached fuel rod tested in argon.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pahari, S.; Hajela, S.; Rammohan, H. P.

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 andmore » 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)« less

Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

NASA Astrophysics Data System (ADS)

Seiler, J. M.; Rameau, B.

Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.

A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11--12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented.

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toman, G.; Gazdzinski, R.; O`Hearn, E.

1994-02-01

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specificmore » aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.« less

Hydrothermal corrosion of silicon carbide joints without radiation

DOE PAGES

Koyanagi, Takaaki; Katoh, Yutai; Terrani, Kurt A.; ...

2016-09-28

In this paper, hydrothermal corrosion of four types of the silicon carbide (SiC) to SiC plate joints were investigated under pressurized water reactor and boiling water reactor relevant chemical conditions without irradiation. The joints were formed by metal diffusion bonding using molybdenum or titanium interlayer, reaction sintering using Ti—Si—C system, and SiC nanopowder sintering. Most of the joints withstood the corrosion tests for five weeks. The recession of the SiC substrates was limited. Based on the recession of the bonding layers, it was concluded that all the joints except for the molybdenum diffusion bond are promising under the reducing environmentsmore » without radiation. Finally, the SiC nanopowder sintered joint was the most corrosion tolerant under the oxidizing environment among the four joints.« less

Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

DOEpatents

Oosterkamp, W.J.; Marquino, W.

1999-01-05

A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

Corrosion fatigue of alloys 600 and 690 in simulated LWR environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruther, W.E.; Soppett, W.K.; Kassner, T.F.

1996-04-01

Crack growth data were obtained on fracture-mechanics specimens of Alloys 600 and 690 to investigate environmentally assisted cracking (EAC) in simulated boiling water reactor and pressurized water reactor environments at 289 and 320 C. Preliminary information was obtained on the effect of temperature, load ratio, stress intensity (K), and the dissolved-oxygen and -hydrogen concentrations of the water on EAC. Specimens of Type 316NG and sensitized Type 304 stainless steel (SS) were included in several of the experiments to assess the behavior of these materials and Alloy 600 under the same water chemistry and loading conditions. The experimental data are comparedmore » with predictions from an Argonne National Laboratory (ANL) model for crack growth rates (CGRs) of SSs in water and the ASME Code Section 11 correlation for CGRs in air at the K{sub max} and load-ratio values in the various tests. The data for all of the materials were bounded by ANL model predictions and the ASME Section 11 ``air line.``« less

Method for depleting BWRs using optimal control rod patterns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1991-01-01

Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonicsmore » calculations.« less

FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

DOEpatents

Treshow, M.

1962-12-25

An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

21 CFR 872.6710 - Boiling water sterilizer.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling water...

21 CFR 872.6710 - Boiling water sterilizer.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling water...

21 CFR 872.6710 - Boiling water sterilizer.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 8 2012-04-01 2012-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling water...

21 CFR 872.6710 - Boiling water sterilizer.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 8 2013-04-01 2013-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling water...

A User’s Guide to the PLTEMP/ANL Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Kalimullah, M.; Feldman, E. E.

2016-07-25

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Graslund, C.; Hellstrand, E.

Sweden benefits in many ways from the reactor safety research performed in other countries. Its own activity complements this effort, but a certain fraction is oriented toward safety issues that are intimately related to the special design of the ASEA-ATOM boiling-water reactor. Through the availability of the decommissioned Marviken reactor plant, Sweden has been able to play a leading role in integral containment experiments with international participation. Joint efforts with other countries are now devoted to defining new large-scale experiments to be performed in the unique Marviken facility. The largest portion of the safety research program in Sweden is performedmore » by Studsvik Energiteknik AB, but various universities, consultant firms, and research institutes are also involved. In addition, a substantial amount of work is done by the reactor vendor ASEA-ATOM. The overall annual budget is at present between $7 and $8 million, with three governmental authorities as the main financing bodies.« less

ALARA Council: Sharing our resources and experiences to reduce doses at Commonwealth Edison Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rescek, F.

1995-03-01

Commonwealth Edison Company is an investor-owned utility company supplying electricity to over three million customers (eight million people) in Chicago and northern Illinois, USA. The company operates 16 generating stations which have the capacity to produce 22,522 megawatts of electricity. Six of these generating stations, containing 12 nuclear units, supply 51% of this capacity. The 12 nuclear units are comprised of four General Electric boiling water (BWR-3) reactors, two General Electric BWR-5 reactors, and six Westinghouse four-loop pressurized water reactors (PWR). In August 1993, Commonwealth Edison created an ALARA Council with the responsibility to provide leadership and guidance that resultsmore » in an effective ALARA Culture within the Nuclear Operations Division. Unlike its predecessor, the Corporate ALARA Committee, the ALARA Council is designed to bring together senior managers from the six nuclear stations and corporate to create a collaborative effort to reduce occupational doses at Commonwealth Edison`s stations.« less

TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra,A.

2013-11-10

A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

New Folklore about Water.

ERIC Educational Resources Information Center

LeMaire, Peter; Waiveris, Charles

1995-01-01

Describes experiments designed to investigate the cooling rate of microwave-boiled water as compared to that of stove-boiled water. Concludes that within experimental limits, microwave-boiled water and stove-boiled water cool at the same rate. (JRH)

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki; Anshari, Rio

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environmentmore » such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.« less

Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

NASA Astrophysics Data System (ADS)

Su'ud, Zaki; Anshari, Rio

2012-06-01

Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Der Sluys, W.A.; Emanuelson, R.H.

1986-01-01

During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

TRAC-BF1 thermal-hydraulic, ANSYS stress analysis for core shroud cracking phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shoop, U.; Feltus, M.A.; Baratta, A.J.

1996-12-31

The U.S. Nuclear Regulatory Commission sent Generic Letter 94-03 informing all licensees about the intergranular stress corrosion cracking (IGSCC) of core shrouds found in both Dresden unit I and Quad Cities unit 1. The letter directed all licensees to perform safety analysis of their boiling water reactor (BWR) units. Two transients of special concern for the core shroud safety analysis include the main steam line break (MSLB) and recirculation line break transient.

Interim reliability evaluation program, Browns Ferry 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, S.E.; Poloski, J.P.; Sullivan, W.H.

1981-01-01

Probabilistic risk analysis techniques, i.e., event tree and fault tree analysis, were utilized to provide a risk assessment of the Browns Ferry Nuclear Plant Unit 1. Browns Ferry 1 is a General Electric boiling water reactor of the BWR 4 product line with a Mark 1 (drywell and torus) containment. Within the guidelines of the IREP Procedure and Schedule Guide, dominant accident sequences that contribute to public health and safety risks were identified and grouped according to release categories.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation,more » boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.« less

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

DOEpatents

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernander, O.; Haga, I.; Segerberg, F.

BS>From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). Although the present status of the boiling water reactor is one of proven technology, design refinements and technical innovations are still being made to further improve reliability, economy and safety. The new standard ASEA- ATOM BWR features a number of such refinements and design improvements involving main circulation punips, containment design, refuelling system and off-gas treatment plant. In some respects the nuclear and hydraulic design of the ASEA- ATOM BWR differs from that adopted by other BWR manufacturers. Since the Oskarshamn I plant was the first nuclear power station havingmore » these features an extensive physics and hydraulics test program was made during the reactor start- up. The results of these tests have fully confirmed the ability of calculation methods to predict the behavior of the reactor. (auth)« less

The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

NASA Astrophysics Data System (ADS)

Cole, Christopher J. P.

Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas that should be investigated. These include developing a detailed point nodel kinetic model coupled with a finite element heat transfer model, undertaking radiation protection shielding calculations in accordance with international and national regulations, and exploring the effects of advanced fuels.

Household Water Disinfection in Hurricane-Affected Communities of Louisiana: Implications for Disaster Preparedness for the General Public

PubMed Central

Ram, Pavani K.; Blanton, Elizabeth; Klinghoffer, Debra; Platek, Mary; Piper, Janet; Straif-Bourgeois, Susanne; Bonner, Matthew R.; Mintz, Eric D.

2007-01-01

Objectives. Thousands of Louisiana residents were asked to boil water because of widespread disruptions in electricity and natural gas services after Hurricane Rita. We sought to assess awareness of boil water orders and familiarity with household water disinfection techniques other than boiling. Methods. We conducted a cross-sectional survey in randomly selected mobile home communities in Louisiana. Results. We interviewed 196 respondents from 8 communities, which had boil water orders instituted. Of 97 who were home while communities were still under orders to boil water, 30 (31%) were aware of the orders and, of those, 24 (80%) said the orders were active while they were living at home; of the 24, 10 (42%) reported boiling water. Overall, 163 (83%) respondents were aware of a method of water disinfection at the household level: boiling (78%), chlorination (27%), and filtration (25%); 87% had a container of chlorine bleach at home. Conclusions. Few hurricane-affected respondents were aware of boil water orders and of alternate water disinfection techniques. Most had access to chlorine and could have practiced household chlorination if disruption in natural gas and electricity made boiling impossible. PMID:17413065

Experimental and numerical investigation of HyperVapotron heat transfer

NASA Astrophysics Data System (ADS)

Wang, Weihua; Deng, Haifei; Huang, Shenghong; Chu, Delin; Yang, Bin; Mei, Luoqin; Pan, Baoguo

2014-12-01

The divertor first wall and neutral beam injection (NBI) components of tokamak devices require high heat flux removal up to 20-30 MW m-2 for future fusion reactors. The water cooled HyperVapotron (HV) structure, which relies on internal grooves or fins and boiling heat transfer to maximize the heat transfer capability, is the most promising candidate. The HV devices, that are able to transfer large amounts of heat (1-20 MW m-2) efficiently, have therefore been developed specifically for this application. Until recently, there have been few attempts to observe the detailed bubble characteristics and vortex evolvement of coolant flowing inside their various parts and understand of the internal two-phase complex heat transfer mechanism behind the vapotron effect. This research builds the experimental facilities of HyperVapotron Loop-I (HVL-I) and Pressure Water HyperVapotron Loop-II (PWHL-II) to implement the subcooled boiling principle experiment in terms of typical flow parameters, geometrical parameters of test section and surface heat flux, which are similar to those of the ITER-like first wall and NBI components (EAST and MAST). The multiphase flow and heat transfer phenomena on the surface of grooves and triangular fins when the subcooled water flowed through were observed and measured with the planar laser induced fluorescence (PLIF) and high-speed photography (HSP) techniques. Particle image velocimetry (PIV) was selected to reveal vortex formation, the flow structure that promotes the vapotron effect during subcooled boiling. The coolant flow data for contributing to the understanding of the vapotron phenomenon and the assessment of how the design and operational conditions that might affect the thermal performance of the devices were collected and analysed. The subcooled flow boiling model and methods of HV heat transfer adopted in the considered computational fluid dynamics (CFD) code were evaluated by comparing the calculated wall temperatures with the experimentally measured values. It was discovered that the bubble and vortex characteristics in the HV are clearly heavily dependent on the internal geometry, flow conditions and input heat flux. The evaporation latent heat is the primary heat transfer mechanism of HV flow under the condition of high heat flux, and the heat transfer through convection is very limited. The percentage of wall heat flux going into vapour production is almost 70%. These relationships between the flow phenomena and thermal performance of the HV device are essential to study the mechanisms for the flow structure alterations for design optimization and improvements of the ITER-like devices' water cooling structure and plasma facing components for future fusion reactors.

Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Y.A.; Feltus, M.A.

1995-07-01

Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specificmore » MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company`s Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates.« less

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

NASA Astrophysics Data System (ADS)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

NASA Astrophysics Data System (ADS)

Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

2016-10-01

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

Uniform corrosion of FeCrAl alloys in LWR coolant environments

DOE PAGES

Terrani, K. A.; Pint, B. A.; Kim, Y. -J.; ...

2016-06-29

The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation ofmore » very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.« less

Trace Assessment for BWR ATWS Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, L.Y.; Diamond, D.; Arantxa Cuadra, Gilad Raitses, Arnold Aronson

2010-04-22

A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtainedmore » from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depres-surization system. The model is not considered complete and recommendations are made on how it should be improved.« less

Disparities in nuclear power plant performance in the United States and the Federal Republic of Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hansen, K.F.; Winje, D.K.

This report presents data comparing the performance of light water reactors in the United States and the Federal Republic of Germany (FRG). The comparisons are made for the years 1980 to 1983 and include 21 Westinghouse Pressurized Water Reactors (PWRs), 22 General Electric Boiling Water Reactors (BWRs) in the US; and 6 Kraftwerk Union (KWU) PWRs and 4 KWU BWRs in the FRG. Data on capacity losses are presented in a disaggregated form for scheduled outages, forced outages, and regulatory imposed outages. Further, within the scheduled and forced outages, the data are subdivided into losses associated with the nuclear island,more » the balance of plant, or other causes. The report also surveys a number of observations relating to the causes of discrepancies between the US and West Germany. These observations were obtained from interviews with executives and engineers in both nations, including people from vendors, utilities, regulators, and architect/engineers. These discussions are distilled into observations relating to national differences in the broad areas of economics and economic regulation, safety regulations, and technical and managerial differences. 9 refs., 2 figs., 26 tabs.« less

Multiphysics modeling of two-phase film boiling within porous corrosion deposits

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Miaomiao, E-mail: mmjin@mit.edu; Short, Michael, E-mail: hereiam@mit.edu

2016-07-01

Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits.more » Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.« less

System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment

NASA Astrophysics Data System (ADS)

Usov, E. V.; Butov, A. A.; Dugarov, G. A.; Kudasov, I. G.; Lezhnin, S. I.; Mosunova, N. A.; Pribaturin, N. A.

2017-07-01

The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.

Gadolinia fuel performance in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bailey, W.E.; Crowther, R.L.

1985-11-01

Gadolinia has the unique property of having a high neutron absorption cross section coupled with a burnup rate that can approximately match the uranium 235 depletion. These qualities and others make gadolinia an ideal burnable absorber, and it has been used in all General Electric-designed boiling water reactors. Fabrication corrosion properties, and performance of gadolinia-containing fuel elements are discussed. Development of a reliable and efficient set of local and global gadolinia-urania design methods has been an arduous process that has taken approx.15 years to accomplish.

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readilymore » achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.« less

Hydrogen considerations in light-water power reactons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keilholtz, G.W.

1976-02-01

A critical review of the literature now available on hydrogen considerations in light-water power reactors (LWRs) and a bibliography of that literature are presented. The subject matter includes mechanisms for the generation of hydrogen-oxygen mixtures, a description of the fundamental properties of such mixtures, and their spontaneous ignition in both static and dynamic systems. The limits for hydrogen flammability and flame propagation are examined in terms of the effects of pressure, temperature, and additives; the emphasis is on the effects of steam and water vapor. The containment systems for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are compared, and methodsmore » to control hydrogen and oxygen under the conditions of both normal operation and postulated accidents are reviewed. It is concluded that hydrogen can be controlled so that serious complications from the production of hydrogen will not occur. The bibliography contains abstracts from the computerized files of the Nuclear Safety Information Center. Key-word, author, and permuted-title indexes are provided. The bibliography includes responses to questions asked by the U. S. Nuclear Regulatory Commission (NRC) which relate to hydrogen, as well as information on normal operations and postulated accidents including generation of hydrogen from core sprays. Other topics included in the ten sections of the bibliography are metal-water reactions, containment atmosphere, radiolytic gas, and recombiners.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

RETRAN analysis of multiple steam generator blow down caused by an auxiliary feedwater steam-line break

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1987-01-01

Analysis results for multiple steam generator blow down caused by an auxiliary feedwater steam-line break performed with the RETRAN-02 MOD 003 computer code are presented to demonstrate the capabilities of the RETRAN code to predict system transient response for verifying changes in operational procedures and supporting plant equipment modifications. A typical four-loop Westinghouse pressurized water reactor was modeled using best-estimate versus worst case licensing assumptions. This paper presents analyses performed to evaluate the necessity of implementing an auxiliary feedwater steam-line isolation modification. RETRAN transient analysis can be used to determine core cooling capability response, departure from nucleate boiling ratio (DNBR)more » status, and reactor trip signal actuation times.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leppik, P.A.

This paper presents results of a study designed to confirm that the interaction of the neutron flux and the coolant flow plays an important role in the mechanism of high-frequency (HF) resonant instability of the VK-50 boiling water reactor. To do this and to check the working model, signals from probes measuring the flow rate of the coolant and the neutron flux were recorded simultaneously (with the help of a magnetograph) in experiments performed in 1981 on driving the VK-50 reactor into the HF reonant instability regimes. Estimates were then obtained for the statistical characteristics of the pulsations of themore » flow rate and of the neutron flux, including the cross-correlation functions and coherence functions. The basic results of these studies are reported here.« less

Calculation of the temperature in the container unit with a modified design for the production of {sup 99}Mo at the VVR-Ts research reactor facility (IVV.10M)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com; Sergeev, V. V.; Kochnov, O. Yu.

The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.

Application of reliability-centered-maintenance to BWR ECCS motor operator valve performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Choi, Y.A.

1993-01-01

This paper describes the application of reliability-centered maintenance (RCM) methods to plant probabilistic risk assessment (PRA) and safety analyses for four boiling water reactor emergency core cooling systems (ECCSs): (1) high-pressure coolant injection (HPCI); (2) reactor core isolation cooling (RCIC); (3) residual heat removal (RHR); and (4) core spray systems. Reliability-centered maintenance is a system function-based technique for improving a preventive maintenance program that is applied on a component basis. Those components that truly affect plant function are identified, and maintenance tasks are focused on preventing their failures. The RCM evaluation establishes the relevant criteria that preserve system function somore » that an RCM-focused approach can be flexible and dynamic.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

Developments in SCC Mitigation by Electrocatalysis

NASA Astrophysics Data System (ADS)

Andresen, Peter L.; Kim, Young J.

SCC is strongly influenced by water chemistry parameters, especially when crack chemistry can be concentrated from differential aeration or thermal gradients or boiling. Mitigation of the effects of the high corrosion potential associated with oxidants is markedly and efficiently accomplished by electrocatalysis, which requires that there be a stoichiometric excess of reductants over oxidants. Mechanisms and criteria for effective SCC mitigation are summarized, with particular focus on the critical location for the catalyst in a crack and experimental support for these concepts. Optimization of electrocatalysis by OnLine NobleChem- is described, for example where Pt is injected at levels of 0.002 to 0.05 ppb in the reactor water.

Parents of two-phase flow and theory of "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper gives a brief overview of types of two-phase flow. Subsequently, it deals with their mutual division and problems with accuracy boundaries among particular types. It also shows the case of water flow through a pipe with external heating and the gradual origination of all kinds of flow. We have met it in solution of safety condition of various stages in pressurized and boiling water reactors. In the MSR there is a problem in the solution of gas-lift using helium as a gas and its secondary usage for clearing of the fuel mixture from gaseous fission products. Theory of gas-lift is described.

Microbiological effectiveness of disinfecting water by boiling in rural Guatemala.

PubMed

Rosa, Ghislaine; Miller, Laura; Clasen, Thomas

2010-03-01

Boiling is the most common means of treating water in the home and the benchmark against which alternative point-of-use water treatment options must be compared. In a 5-week study in rural Guatemala among 45 households who claimed they always or almost always boiled their drinking water, boiling was associated with a 86.2% reduction in geometric mean thermotolerant coliforms (TTC) (N = 206, P < 0.0001). Despite consistent levels of fecal contamination in source water, 71.2% of stored water samples from self-reported boilers met the World Health Organization guidelines for safe drinking water (0 TTC/100 mL), and 10.7% fell within the commonly accepted low-risk category of (1-10 TTC/100 mL). As actually practiced in the study community, boiling significantly improved the microbiological quality of drinking water, though boiled and stored drinking water is not always free of fecal contaminations.

Microbiological Effectiveness of Disinfecting Water by Boiling in Rural Guatemala

PubMed Central

Rosa, Ghislaine; Miller, Laura; Clasen, Thomas

2010-01-01

Boiling is the most common means of treating water in the home and the benchmark against which alternative point-of-use water treatment options must be compared. In a 5-week study in rural Guatemala among 45 households who claimed they always or almost always boiled their drinking water, boiling was associated with a 86.2% reduction in geometric mean thermotolerant coliforms (TTC) (N = 206, P < 0.0001). Despite consistent levels of fecal contamination in source water, 71.2% of stored water samples from self-reported boilers met the World Health Organization guidelines for safe drinking water (0 TTC/100 mL), and 10.7% fell within the commonly accepted low-risk category of (1–10 TTC/100 mL). As actually practiced in the study community, boiling significantly improved the microbiological quality of drinking water, though boiled and stored drinking water is not always free of fecal contaminations. PMID:20207876

When water does not boil at the boiling point.

PubMed

Chang, Hasok

2007-03-01

Every schoolchild learns that, under standard pressure, pure water always boils at 100 degrees C. Except that it does not. By the late 18th century, pioneering scientists had already discovered great variations in the boiling temperature of water under fixed pressure. So, why have most of us been taught that the boiling point of water is constant? And, if it is not constant, how can it be used as a 'fixed point' for the calibration of thermometers? History of science has the answers.

Environmental Assessment for Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear Superheat (BONUS) Reactor Building, Rincon, Puerto Rico

DOE Office of Scientific and Technical Information (OSTI.GOV)

N /A

The U.S. Department of Energy (DOE) proposes to consent to a proposal by the Puerto Rico Electric Power Authority (PREPA) to allow public access to the Boiling Nuclear Superheat (BONUS) reactor building located near Rincon, Puerto Rico for use as a museum. PREPA, the owner of the BONUS facility, has determined that the historical significance of this facility, as one of only two reactors of this design ever constructed in the world, warrants preservation in a museum, and that this museum would provide economic benefits to the local community through increased tourism. Therefore, PREPA is proposing development of the BONUSmore » facility as a museum.« less

Electric kettles as a source of human lead exposure.

PubMed

Wigle, D T; Charlebois, E J

1978-01-01

Five hundred and seventy-four households in Ottawa were surveyed to evaluate water boiled in electric kettles as a source of lead exposure. Samples of boiled water exceeded the World Health Organization mandatory limit for drinking water (50 microgram/l) in 42.5% of the households. Excessive lead concentrations were observed in 62.8% of water samples from kettles more than 5 years old. Multiple regression analysis indicated that age, sex, and cigarette smoking habits, but not lead concentration in boiled water, nor weekly consumption of boiled water were significantly associated with blood-lead concentration. Lead exposure from electric kettles may be a significant problem only in infants receiving formula prepared with boiled water.

Incorporating Water Boiling in the Numerical Modelling of Thermal Remediation by Electrical Resistance Heating

NASA Astrophysics Data System (ADS)

Molnar, I. L.; Krol, M.; Mumford, K. G.

2017-12-01

Developing numerical models for subsurface thermal remediation techniques - such as Electrical Resistive Heating (ERH) - that include multiphase processes such as in-situ water boiling, gas production and recovery has remained a significant challenge. These subsurface gas generation and recovery processes are driven by physical phenomena such as discrete and unstable gas (bubble) flow as well as water-gas phase mass transfer rates during bubble flow. Traditional approaches to multiphase flow modeling soil remain unable to accurately describe these phenomena. However, it has been demonstrated that Macroscopic Invasion Percolation (MIP) can successfully simulate discrete and unstable gas transport1. This has lead to the development of a coupled Electro Thermal-MIP Model2 (ET-MIP) capable of simulating multiple key processes in the thermal remediation and gas recovery process including: electrical heating of soil and groundwater, water flow, geological heterogeneity, heating-induced buoyant flow, water boiling, gas bubble generation and mobilization, contaminant mass transport and removal, and additional mechanisms such as bubble collapse in cooler regions. This study presents the first rigorous validation of a coupled ET-MIP model against two-dimensional water boiling and water/NAPL co-boiling experiments3. Once validated, the model was used to explore the impact of water and co-boiling events and subsequent gas generation and mobilization on ERH's ability to 1) generate, expand and mobilize gas at boiling and NAPL co-boiling temperatures, 2) efficiently strip contaminants from soil during both boiling and co-boiling. In addition, a quantification of the energy losses arising from steam generation during subsurface water boiling was examined with respect to its impact on the efficacy of thermal remediation. While this study specifically targets ERH, the study's focus on examining the fundamental mechanisms driving thermal remediation (e.g., water boiling) renders these results applicable to a wide range of thermal and gas-based remediation techniques. 1. Mumford, K. G., et al. (2010), Adv. Water Resour. 2010, 33 (4), 504-513. 2. Krol, M. M., et al. (2011), Adv. Water Resour. 2011, 34 (4), 537-549. 3. Hegele, P. R. and Mumford, K. G. Journal of Contaminant Hydrology 2014, 165, 24-36.

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Can we remove iodine-131 from tap water in Japan by boiling? - Experimental testing in response to the Fukushima Daiichi Nuclear Power Plant accident.

PubMed

Tagami, K; Uchida, S

2011-08-01

Iodine-131 concentrations in tap water higher than 100 BqL(-1) were reported by several local governments in Japan following the Fukushima Daiichi Nuclear Power Plant accident. Some individuals in the emergency-response community recommended the boiling of tap water to remove iodine-131. However, the tap water boiling tests in this study showed no iodine-131 loss from the tap water with either short-term boiling (1-10 min) or prolonged boiling (up to 30 min) resulting in up to 3-fold volume reductions. In this situation, boiling was shown to be not effective in removing iodine-131 from tap water; indeed even higher concentrations may result from the liquid-volume reduction accompanying this process. Copyright © 2011 Elsevier Ltd. All rights reserved.

Standard-less analysis of Zircaloy clad samples by an instrumental neutron activation method

NASA Astrophysics Data System (ADS)

Acharya, R.; Nair, A. G. C.; Reddy, A. V. R.; Goswami, A.

2004-03-01

A non-destructive method for analysis of irregular shape and size samples of Zircaloy has been developed using the recently standardized k0-based internal mono standard instrumental neutron activation analysis (INAA). The samples of Zircaloy-2 and -4 tubes, used as fuel cladding in Indian boiling water reactors (BWR) and pressurized heavy water reactors (PHWR), respectively, have been analyzed. Samples weighing in the range of a few tens of grams were irradiated in the thermal column of Apsara reactor to minimize neutron flux perturbations and high radiation dose. The method utilizes in situ relative detection efficiency using the γ-rays of selected activation products in the sample for overcoming γ-ray self-attenuation. Since the major and minor constituents (Zr, Sn, Fe, Cr and/or Ni) in these samples were amenable to NAA, the absolute concentrations of all the elements were determined using mass balance instead of using the concentration of the internal mono standard. Concentrations were also determined in a smaller size Zircaloy-4 sample by irradiating in the core position of the reactor to validate the present methodology. The results were compared with literature specifications and were found to be satisfactory. Values of sensitivities and detection limits have been evaluated for the elements analyzed.

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Final Summary Design Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1962-05-01

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fedyakin, R.E.

Translated from Teploenergetika; 20: No. 2, 23-24(1973)The water velocity at the inlet of the fuel channels of the VK-50 boiling reactor was recorded to be 0.8 to 0.85 m/s in most cases. Two of the velocity meters indicated a depressed velocity. Checking involved measuring the relative energy release distribution and of the cadmium ratio in different fuel channels. The method is described, and it was concluded that the abnormal deviation shown by the meters indicates a flaw in the sensing instruments, but impossible for replacement because of the inaccessible position of the velocity meters. (MCW)

System Study: High-Pressure Coolant Injection 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

System Study: High-Pressure Core Spray 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

Impact of nonabsorbing control rod tips on kinetics feedback for BWR turbine trip without bypass RETRAN-03 analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Knerr, R.; Shoop, U.

1993-01-01

RETRAN-03 studies were performed for the boiling water reactor (BWR) turbine trip without bypass (TTWOB) event to investigate how the non-neutron-absorbing material on control rod tips affect scram delay timing and reactivity feedback. Scram delay, Doppler temperature, and moderator void (density) feedback were varied to assess their relative impact on kinetics behavior. Although a generic point-kinetics RETRAN-03 TTWOB model 2 was employed, actual plant information was used to develop the basic and parametric cases.

The myth of the boiling point.

PubMed

Chang, Hasok

2008-01-01

Around 1800, many reputable scientists reported significant variations in the temperature of pure water boiling under normal atmospheric pressure. The reported variations included a difference of over 1 degree C between boiling in metallic and glass vessels (Gay-Lussac), and "superheating" up to 112 degrees C on extracting dissolved air out of water (De Luc). I have confirmed most of these observations in my own experiments, many of which are described in this paper. Water boils at the "boiling point" only under very particular circumstances. Our common-sense intuition about the fixedness of the boiling point is only sustained by our limited experience.

Predictors of Drinking Water Boiling and Bottled Water Consumption in Rural China: A Hierarchical Modeling Approach.

PubMed

Cohen, Alasdair; Zhang, Qi; Luo, Qing; Tao, Yong; Colford, John M; Ray, Isha

2017-06-20

Approximately two billion people drink unsafe water. Boiling is the most commonly used household water treatment (HWT) method globally and in China. HWT can make water safer, but sustained adoption is rare and bottled water consumption is growing. To successfully promote HWT, an understanding of associated socioeconomic factors is critical. We collected survey data and water samples from 450 rural households in Guangxi Province, China. Covariates were grouped into blocks to hierarchically construct modified Poisson models and estimate risk ratios (RR) associated with boiling methods, bottled water, and untreated water. Female-headed households were most likely to boil (RR = 1.36, p < 0.01), and among boilers those using electric kettles rather than pots had higher income proxies (e.g., per capita TV ownership RR = 1.42, p < 0.01). Higher-income households with younger, literate, and male heads were more likely to purchase (frequently contaminated) bottled water, or use electric kettles if they boiled. Our findings show that boiling is not an undifferentiated practice, but one with different methods of varying effectiveness, environmental impact, and adoption across socioeconomic strata. Our results can inform programs to promote safer and more efficient boiling using electric kettles, and suggest that if rural China's economy continues to grow then bottled water use will increase.

ATOMIC POWER PLANT

DOEpatents

Daniels, F.

1957-11-01

This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Radiological impact of airborne effluents of coal and nuclear plants.

PubMed

McBride, J P; Moore, R E; Witherspoon, J P; Blanco, R E

1978-12-08

Radiation doses from airborne effluents of model coal-fired and nuclear power plants (1000 megawatts electric) are compared. Assuming a 1 percent ash release to the atmosphere (Environmental Protection Agency regulation) and 1 part per million of uranium and 2 parts per million of thorium in the coal (approximately the U.S. average), population doses from the coal plant are typically higher than those from pressurized-water or boiling-water reactors that meet government regulations. Higher radionuclide contents and ash releases are common and would result in increased doses from the coal plant. The study does not assess the impact of non-radiological pollutants or the total radiological impacts of a coal versus a nuclear economy.

Experience with chemical system decontamination by the CORD process and electrochemical decontamination of pipe ends

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wille, H.; Bertholdt, H.O.; Operschall, H.

Efforts to reduce occupational radiation exposure during inspection and repair work in nuclear power plants turns steadily increasing attention to the decontamination of systems and components. Due to the advanced age of nuclear power plants resulting in increasing dose rates, the decontamination of components, or rather of complete systems, or loops to protect operating and inspection personnel becomes demanding. Besides, decontaminating complete primary loops is in many cases less difficult than cleaning large components. Based on experience gained in nuclear power plants, an outline of two different decontamination methods performed recently are given. For the decontamination of complete systems ormore » loops, Kraftwerk Union AG has developed CORD, a low-concentration process. For the decontamination performance of a subsystem, such as the steam generator (SG) channel heads of a pressurized water reactor or the recirculation loops of a boiling water reactor the automated mobile decontamination appliance is used. The electrochemical decontamination process is primarily applicable for the treatment of specially limited surface areas.« less

Assessing the microbiological performance and potential cost of boiling drinking water in urban Zambia.

PubMed

Psutka, Rebecca; Peletz, Rachel; Michelo, Sandford; Kelly, Paul; Clasen, Thomas

2011-07-15

Boiling is the most common method of disinfecting water in the home and the benchmark against which other point-of-use water treatment is measured. In a six-week study in peri-urban Zambia, we assessed the microbiological effectiveness and potential cost of boiling among 49 households without a water connection who reported "always" or "almost always" boiling their water before drinking it. Source and household drinking water samples were compared weekly for thermotolerant coliforms (TTC), an indicator of fecal contamination. Demographics, costs, and other information were collected through surveys and structured observations. Drinking water samples taken at the household (geometric mean 7.2 TTC/100 mL, 95% CI, 5.4-9.7) were actually worse in microbiological quality than source water (geometric mean 4.0 TTC/100 mL, 95% CI, 3.1-5.1) (p < 0.001), although both are relatively low levels of contamination. Only 60% of drinking water samples were reported to have actually been boiled at the time of collection from the home, suggesting over-reporting and inconsistent compliance. However, these samples were of no higher microbiological quality. Evidence suggests that water quality deteriorated after boiling due to lack of residual protection and unsafe storage and handling. The potential cost of fuel or electricity for boiling was estimated at 5% and 7% of income, respectively. In this setting where microbiological water quality was relatively good at the source, safe-storage practices that minimize recontamination may be more effective in managing the risk of disease from drinking water at a fraction of the cost of boiling.

Fast reactor safety and related physics. Volume IV. Phenomenology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-01-01

Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Relative stress corrosion susceptibilities of alloys 690 and 600 in simulated boiling water reactor environments

NASA Astrophysics Data System (ADS)

Page, R. A.; McMinn, A.

1986-05-01

The relative susceptibilities of alloys 600 and 690 to intergranular stress corrosion cracking (IGSCC) in pure water and a simulated resin intrusion environment at 288 °C were evaluated. A combination of creviced and noncreviced slow-strain-rate, and precracked fracture mechanics tests were employed in the evaluation. Susceptibility was determined as a function of dissolved oxygen content, degree of sensitization, and crevice condition. The results indicated that alloy 600 was susceptible to various degrees of IGSCC in oxygen containing pure water when creviced, and immune to IGSCC when uncreviced. Alloy 690 was immune to IGSCC under all pure water conditions examined. Alloy 600 and alloy 690 were both susceptible to cracking in the simulated resin intrusion environment. Alloy 690, however, exhibited the greatest resistance to SCC of the two alloys.

A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. L. Chichester; S. J. Thompson

2013-09-01

This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium inmore » the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for using cerium, which is rather easy to analyze using passive nondestructive analysis gamma-ray spectrometry, as a surrogate for plutonium in the final analysis of TMI-2 melted fuel debris. The generation of this report is motivated by the need to perform nuclear material accountancy measurements on the melted fuel debris that will be excavated from the damaged nuclear reactors at the Fukushima Daiichi nuclear power plant in Japan, which were destroyed by the Tohoku earthquake and tsunami on March 11, 2011. Lessons may be taken from prior U.S. work related to the study of the TMI-2 core debris to support the development of new assay methods for use at Fukushima Daiichi. While significant differences exist between the two reactor systems (pressurized water reactor (TMI-2) versus boiling water reactor (FD), fresh water post-accident cooing (TMI-2) versus salt water (FD), maintained containment (TMI-2) versus loss of containment (FD)) there remain sufficient similarities to motivate these comparisons.« less

Method of controlling crystallite size in nuclear-reactor fuels

DOEpatents

Lloyd, Milton H.; Collins, Jack L.; Shell, Sam E.

1985-01-01

Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

Method of controlling crystallite size in nuclear-reactor fuels

DOEpatents

Lloyd, M.H.; Collins, J.L.; Shell, S.E.

Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Fatigue crack growth in SA508-CL2 steel in a high temperature, high purity water environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerber, T.L.; Heald, J.D.; Kiss, E.

1974-10-01

Fatigue crack growth tests were conducted with 1 in. plate specimens of SA508-CL 2 steel in room temperature air, 550$sup 0$F air and in a 550$sup 0$F, high purity, water environment. Zero-tension load controlled tests were run at cyclic frequencies as low as 0.037 CPM. Results show that growth rates in the simulated Boiling Water Reactor (BWR) water environment are faster than growth rates observed in 550$sup 0$F air and these rates are faster than the room temperature rate. In the BWR water environment, lowering the cyclic frequency from 0.37 to 0.037 CPM caused only a slight increase in themore » fatigue crack growth rate. All growth rates measured in these tests were below the upper bound design curve presented in Section XI of the ASME Code. (auth)« less

Water boiling inside carbon nanotubes: toward efficient drug release.

PubMed

Chaban, Vitaly V; Prezhdo, Oleg V

2011-07-26

We show using molecular dynamics simulation that spatial confinement of water inside carbon nanotubes (CNTs) substantially increases its boiling temperature and that a small temperature growth above the boiling point dramatically raises the inside pressure. Capillary theory successfully predicts the boiling point elevation down to 2 nm, below which large deviations between the theory and atomistic simulation take place. Water behaves qualitatively different inside narrow CNTs, exhibiting transition into an unusual phase, where pressure is gas-like and grows linearly with temperature, while the diffusion constant is temperature-independent. Precise control over boiling by CNT diameter, together with the rapid growth of inside pressure above the boiling point, suggests a novel drug delivery protocol. Polar drug molecules are packaged inside CNTs; the latter are delivered into living tissues and heated by laser. Solvent boiling facilitates drug release.

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Ultra-High Temperature Steam Corrosion of Complex Silicates for Nuclear Applications: A Computational Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rashkeev, Sergey N.; Glazoff, Michael V.; Tokuhiro, Akira

2014-01-01

Stability of materials under extreme conditions is an important issue for safety of nuclear reactors. Presently, silicon carbide (SiC) is being studied as a cladding material candidate for fuel rods in boiling-water and pressurized water-cooled reactors (BWRs and PWRs) that would substitute or modify traditional zircaloy materials. The rate of corrosion of the SiC ceramics in hot vapor environment (up to 2200 degrees C) simulating emergency conditions of light water reactor (LWR) depends on many environmental factors such as pressure, temperature, viscosity, and surface quality. Using the paralinear oxidation theory developed for ceramics in the combustion reactor environment, we estimatedmore » the corrosion rate of SiC ceramics under the conditions representing a significant power excursion in a LWR. It was established that a significant time – at least 100 h – is required for a typical SiC braiding to significantly degrade even in the most aggressive vapor environment (with temperatures up to 2200 °C) which is possible in a LWR at emergency condition. This provides evidence in favor of using the SiC coatings/braidings for additional protection of nuclear reactor rods against off-normal material degradation during power excursions or LOCA incidents. Additionally, we discuss possibilities of using other silica based ceramics in order to find materials with even higher corrosion resistance than SiC. In particular, we found that zircon (ZrSiO4) is also a very promising material for nuclear applications. Thermodynamic and first-principles atomic-scale calculations provide evidence of zircon thermodynamic stability in aggressive environments at least up to 1535 degrees C.« less

Emergency Disinfection of Drinking Water

EPA Pesticide Factsheets

How to boil and disinfect water to kill most disease-causing microorganisms during emergency situations where regular water service has been interrupted and local authorities recommend using only bottled water, boiled water, or disinfected water.

Single-side conduction modeling for high heat flux coolant channels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyd, R.D. Sr.

In the development of plasma-facing components (PFCs), most investigators have erroneously postulated negligible water critical heat flux dependence on the coolant channel length-to-diameter (L/D) ratio above a constant value of L/D. Although encouraging results have been obtained in characterizing peaking factors for local two-dimensional boiling curves and critical heat flux, additional experimental data and theoretical model development are needed to validate the applicability to PFCs. Both these and related issues will affect the flow boiling correlation and data reduction associated with the development of PFCs for fusion reactors and other physical problems that are dependent on conduction modeling in themore » heat flux spectrum of applications. Both exact solutions and numerical conjugate analyses are presented for a one-side heated (OSH) geometry. The results show (a) the coexistence of three flow regimes inside an OSH circular geometry, (b) the correlational dependence of the inside wall heat flux and temperature, and (c) inaccuracies that could arise in some data reduction procedures.« less

Material distribution in light water reactor-type bundles tested under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Noack, V.; Hagen, S.J.L.; Hofmann, P.

1997-02-01

Severe fuel damage experiments simulating small-break loss-of-coolant accidents have been carried out in the CORA out-of-pile test facility at Forschungszentrum Karlsruhe. Rod bundles with electrically heated fuel rod simulators containing annular UO{sub 2} pellets, UO{sub 2} full pellet rods, and absorber rods of two kinds (Ag/In/Cd to represent pressurized water reactor conditions and B{sub 4}C to represent boiling water reactor and VVER-1000 fuel elements) were subjected to temperature transients up to 2,300 K. A special method was applied to determine the axial mass distribution of bundle materials. The low-temperature melt formation by various interactions between zirconium and components of absorbermore » and spacer grids strongly influences the bundle degradation and material relocation. Absorber materials can separate from the fuel by a noncoherent relocation of the materials at different temperatures. The distributions of solidified materials in the different test bundles show a clear dependence on the axial temperature profile. Coolant channel blockages are observed mainly at the lower end of the bundle, i.e., near the lowest elevation at which an oxidation excursion resulting from the highly exothermic zirconium-steam reaction had been experienced. This elevation corresponds with a steep axial temperature gradient in the maximum temperature attained. Oxide layers on Zircaloy result in reduced melt formation.« less

Issue a Boil-Water Advisory or Wait for Definitive Information? A Decision Analysis

PubMed Central

Wagner, Michael M.; Wallstrom, Garrick L.; Onisko, Agnieszka

2005-01-01

Objective Study the decision to issue a boil-water advisory in response to a spike in sales of diarrhea remedies or wait 72 hours for the results of definitive testing of water and people. Methods Decision analysis. Results In the base-case analysis, the optimal decision is test-and-wait. If the cost of issuing a boil-water advisory is less than 13.92 cents per person per day, the optimal decision is to issue the boil-water advisory immediately. Conclusions Decisions based on surveillance data that are suggestive but not conclusive about the existence of a disease outbreak can be modeled. PMID:16779145

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Characterization of 14C in Swedish light water reactors.

PubMed

Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

2008-08-01

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anderson, Michael T.; Simonen, Fredric A.; Muscara, Joseph

2016-09-01

An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which themore » effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.« less

Fecal contamination of drinking water within peri-urban households, Lima, Peru.

PubMed

Oswald, William E; Lescano, Andrés G; Bern, Caryn; Calderon, Maritza M; Cabrera, Lilia; Gilman, Robert H

2007-10-01

We assessed fecal contamination of drinking water in households in 2 peri-urban communities of Lima, Peru. We measured Escherichia coli counts in municipal source water and, within households, water from principal storage containers, stored boiled drinking water, and water in a serving cup. Source water was microbiologically clean, but 26 (28%) of 93 samples of water stored for cooking had fecal contamination. Twenty-seven (30%) of 91 stored boiled drinking water samples grew E. coli. Boiled water was more frequently contaminated when served in a drinking cup than when stored (P < 0.01). Post-source contamination increased successively through the steps of usage from source water to the point of consumption. Boiling failed to ensure safe drinking water at the point of consumption because of easily contaminated containers and poor domestic hygiene. Hygiene education, better point-of-use treatment and storage options, and in-house water connections are urgently needed.

Microbiological effectiveness and cost of boiling to disinfect drinking water in rural Vietnam.

PubMed

Clasen, Thomas F; Thao, Do Hoang; Boisson, Sophie; Shipin, Oleg

2008-06-15

Despite certain shortcomings, boiling is still the most common means of treating water in the home and the benchmark against which alternative household-based disinfection and filtration methods must be measured. We assessed the microbiological effectiveness and cost of boiling among a vulnerable population relying on unimproved water sources and commonly practicing boiling as a means of disinfecting water. In a 12 week study among 50 households from a rural community in Vietnam, boiling was associated with a 97% reduction in geometric mean thermotolerant coliforms (TTCs) (p < 0.001). Despite high levels of faecal contamination in source water, 37% of stored water samples from self-reported boilers met the WHO standard for safe drinking water (0 TTC/100 mL), and 38.3% fell within the low risk category (1--10 TTC/100 mL). Nevertheless, 60.5% of stored drinking water samples were positive for TTC, with 22.2% falling into the medium risk category (11--100 TTC/100 mL). The estimated cost of wood used to boil water was US$ 0.272 per month for wood collectors and US$ 1.68 per month for wood purchasers, representing approximately 0.48% to 1.04%, respectively, of the average monthly income of participating households.

Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog

The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to addressmore » this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.« less

Occurrence and potential causes of androgenic activities in source and drinking water in China.

PubMed

Hu, Xinxin; Shi, Wei; Wei, Si; Zhang, Xiaowei; Feng, Jianfang; Hu, Guanjiu; Chen, Sulan; Giesy, John P; Yu, Hongxia

2013-09-17

The increased incidences of disorders of male reproductive tract as well as testicular and prostate cancers have been attributed to androgenic pollutants in the environment. Drinking water is one pathway of exposure through which humans can be exposed. In this study, both potencies of androgen receptor (AR) agonists and antagonists were determined in organic extracts of raw source water as well as finished water from waterworks, tap water, boiled water, and poured boiled water in eastern China. Ten of 13 samples of source water exhibited detectable AR antagonistic potencies with AR antagonist equivalents (Ant-AR-EQs) ranging from <15.3 (detection limit) to 140 μg flutamide/L. However, no AR agonistic activity was detected in any source water. All finished water from waterworks, tap water, boiled water, and poured boiled water exhibited neither AR agonistic nor antagonistic activity. Although potential risks are posed by source water, water treatment processes effectively removed AR antagonists. Boiling and pouring of water further removed these pollutants. Phthalate esters (PAEs) including diisobutyl phthalate (DIBP) and dibutyl phthalate (DBP) were identified as major contributors to AR antagonistic potencies in source waters. Metabolites of PAEs exhibited no AR antagonistic activity and did not increase potencies of PAEs when they coexist.

Boiling local heat transfer enhancement in minichannels using nanofluids

PubMed Central

2013-01-01

This paper reports an experimental study on nanofluid convective boiling heat transfer in parallel rectangular minichannels of 800 μm hydraulic diameter. Experiments are conducted with pure water and silver nanoparticles suspended in water base fluid. Two small volume fractions of silver nanoparticles suspended in water are tested: 0.000237% and 0.000475%. The experimental results show that the local heat transfer coefficient, local heat flux, and local wall temperature are affected by silver nanoparticle concentration in water base fluid. In addition, different correlations established for boiling flow heat transfer in minichannels or macrochannels are evaluated. It is found that the correlation of Kandlikar and Balasubramanian is the closest to the water boiling heat transfer results. The boiling local heat transfer enhancement by adding silver nanoparticles in base fluid is not uniform along the channel flow. Better performances and highest effect of nanoparticle concentration on the heat transfer are obtained at the minichannels entrance. PMID:23506445

Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

Passive gamma analysis of the boiling-water-reactor assemblies

NASA Astrophysics Data System (ADS)

Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

Passive gamma analysis of the boiling-water-reactor assemblies

DOE PAGES

Vo, D.; Favalli, A.; Grogan, B.; ...

2016-09-01

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R

2015-01-01

Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditionalmore » Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.« less

Sinus Rinsing and Neti Pots

MedlinePlus

... that it contains distilled or sterile water. 2. Filter (if boiled, sterile, or distilled water not available) ... if no boiled or distilled water and no filter available) Disinfect: Learn how to disinfect your water ...

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-05-01

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Ground Water Rule - Boil Water Advisory - Public Notification Template

EPA Pesticide Factsheets

The Ground Water Rule - Boil Water Advisory - Public Notification Template can be use to issue a Tier 1 Public Notification when it has been determined that source ground water is contaminated with E. Coli bacteria.

ANALYSIS OF BORON DILUTION TRANSIENTS IN PWRS.

DOE Office of Scientific and Technical Information (OSTI.GOV)

DIAMOND,D.J.BROMLEY,B.P.ARONSON,A.L.

2004-02-04

A study has been carried out with PARCS/RELAP5 to understand the consequences of hypothetical boron dilution events in pressurized water reactors. The scenarios of concern start with a small-break loss-of-coolant accident. If the event leads to boiling in the core and then the loss of natural circulation, a boron-free condensate can accumulate in the cold leg. The dilution event happens when natural circulation is re-established or a reactor coolant pump (RCP) is restarted in violation of operating procedures. This event is of particular concern in B&W reactors with a lowered-loop design and is a Generic Safety Issue for the U.S.more » Nuclear Regulatory Commission. The results of calculations with the reestablishment of natural circulation show that there is no unacceptable fuel damage. This is determined by calculating the maximum fuel pellet enthalpy, based on the three-dimensional model, and comparing it with the criterion for damage. The calculation is based on a model of a B&W reactor at beginning of the fuel cycle. If an RCP is restarted, unacceptable fuel damage may be possible in plants with sufficiently large volumes of boron-free condensate in the cold leg.« less

Modelling the activity of 129I in the primary coolant of a CANDU reactor

NASA Astrophysics Data System (ADS)

Lewis, Brent J.; Husain, Aamir

2003-01-01

A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 -9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ˜2×10 -8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

Stability Estimation of ABWR on the Basis of Noise Analysis

NASA Astrophysics Data System (ADS)

Furuya, Masahiro; Fukahori, Takanori; Mizokami, Shinya; Yokoya, Jun

In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel stability and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUS-F facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A real-time simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying auto regressive (AR) methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

Benefits of barrier fuel on fuel cycle economics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowther, R.L.; Kunz, C.L.

1988-01-01

Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect ofmore » fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs.« less

Steel Containment Vessel Model Test: Results and Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Hessheimer, M.F.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformationmore » behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.« less

Modeling of the processes of natural and waste water purification in the reactor-clarifier

NASA Astrophysics Data System (ADS)

Primak, O. D.; Skolubovich, Yu. L.; Fedorova, N. N.; Voitov, E. L.

2018-03-01

The results of the filtration process simulation in a reactor-clarifier installation using a suspended loading layer are presented. Calculations were carried out in ANSYS Fluent on the basis of the Navier-Stokes equations supplemented by the equations of the Eulerian model of multiphase taking into account granularity of the particle phase. The unsteady picture of the formation of a fluidized («boiling») layer of particles is obtained. The results of parametric calculations allowing to estimate the effect of the flow velocity, the loading layer thickness, the thickness of sand and other parameters on the fluidized bed structure are presented. The liquid flow rate at which the loading grains are not washed out is determined. The diameter of particles and the height of the loading layer, at which the filter material is suspended and thus normal operation of the plant is ensured, are defined.

Performance of the electronic personal dosemeter for neutron 'Saphydose-N' at different workplaces of nuclear facilities.

PubMed

Lahaye, T; Chau, Q; Ménard, S; Lacoste, V; Muller, H; Luszik-Bhadra, M; Reginatto, M; Bruguier, P

2006-01-01

This paper mainly aims at presenting the measurements and the results obtained with the electronic personal neutron dosemeter Saphydose-N at different facilities. Three campaigns were led in the frame of the European contract EVIDOS ('Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields'). The first one consisted in the measurements at the IRSN French research laboratory in reference neutron fields generated by a thermal facility (SIGMA), radionuclide ISO sources ((241)AmBe; (252)Cf; (252)Cf(D(2)O)\\Cd) and a realistic spectrum (CANEL/T400). The second one was performed at the Krümmel Nuclear Power Plant (Germany) close to the boiling water reactor and to a spent fuel transport cask. The third one was realised at Mol (Belgium), at the VENUS Research Reactor and at Belgonucléaire, a fuel processing factory.

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

Thermionic converter temperature controller

DOEpatents

Shaner, Benjamin J [McMurray, PA; Wolf, Joseph H [Pittsburgh, PA; Johnson, Robert G. R. [Trafford, PA

2001-04-24

A method and apparatus for controlling the temperature of a thermionic reactor over a wide range of operating power, including a thermionic reactor having a plurality of integral cesium reservoirs, a honeycomb material disposed about the reactor which has a plurality of separated cavities, a solid sheath disposed about the honeycomb material and having an opening therein communicating with the honeycomb material and cavities thereof, and a shell disposed about the sheath for creating a coolant annulus therewith so that the coolant in the annulus may fill the cavities and permit nucleate boiling during the operation of the reactor.

Folate content and retention in commonly consumed vegetables in the South Pacific.

PubMed

Maharaj, Prayna P P; Prasad, Surendra; Devi, Riteshma; Gopalan, Romila

2015-09-01

This paper reports the effect of boiling and frying on the retention of folate in commonly consumed Fijian vegetables (drumstick leaves, taro leaves, bele leaves, amaranth leaves, fern/ota, okra and French bean). The folate content was determined by microbiological assay (Lactobacillus casei rhamnosus) and tri-enzyme (protease, α-amylase and chicken pancreas conjugase) extraction treatment. The folate loss varied among the vegetables from 10-64% on boiling while 1-36% on frying. The higher folate loss was observed during boiling. The folate content in the water derived after boiling different vegetables ranged from 11.9 ± 0.5 to 61.6 ± 2.5 μg/100mL. The folate loss on boiling was accounted for in the cooking water. The predominant way of folate loss on boiling was leaching rather than thermal degradation which makes boiling the better choice of cooking the studied vegetables for folate intake, provided the cooking water is consumed together with the vegetables. Copyright © 2015 Elsevier Ltd. All rights reserved.

Catalytic multi-stage process for hydroconversion and refining hydrocarbon feeds

DOEpatents

Comolli, Alfred G.; Lee, Lap-Keung

2001-01-01

A multi-stage catalytic hydrogenation and hydroconversion process for heavy hydrocarbon feed materials such as coal, heavy petroleum fractions, and plastic waste materials. In the process, the feedstock is reacted in a first-stage, back-mixed catalytic reactor with a highly dispersed iron-based catalyst having a powder, gel or liquid form. The reactor effluent is pressure-reduced, vapors and light distillate fractions are removed overhead, and the heavier liquid fraction is fed to a second stage back-mixed catalytic reactor. The first and second stage catalytic reactors are operated at 700-850.degree. F. temperature, 1000-3500 psig hydrogen partial pressure and 20-80 lb./hr per ft.sup.3 reactor space velocity. The vapor and light distillates liquid fractions removed from both the first and second stage reactor effluent streams are combined and passed to an in-line, fixed-bed catalytic hydrotreater for heteroatom removal and for producing high quality naphtha and mid-distillate or a full-range distillate product. The remaining separator bottoms liquid fractions are distilled at successive atmospheric and vacuum pressures, low and intermediate-boiling hydrocarbon liquid products are withdrawn, and heavier distillate fractions are recycled and further upgraded to provide additional low-boiling hydrocarbon liquid products. This catalytic multistage hydrogenation process provides improved flexibility for hydroprocessing the various carbonaceous feedstocks and adjusting to desired product structures and for improved economy of operations.

Characterizing preferential groundwater discharge through boils using temperature

NASA Astrophysics Data System (ADS)

Vandenbohede, A.; de Louw, P. G. B.; Doornenbal, P. J.

2014-03-01

In The Netherlands, preferential groundwater discharge trough boils is a key process in the salinization of deep polders. Previous work showed that boils also influence the temperature in the subsurface and of surface water. This paper elaborates on this process combining field observations with numerical modeling. As is the case for salinity, a distinct anomaly in the subsurface and surface water temperature can be attributed to boils. Lines of equal temperature are distorted towards the boil, which can be considered as an upconing of the temperature profile by analogy of the upconing of a fresh-saltwater interface. The zone of this distortion is limited to the immediate vicinity of the boil, being about 5 m in the aquitard which holds the boil's conduit, or maximum a few dozens of meters in the underlying aquifer. In the aquitard, heat transport is conduction dominated whereas this is convection dominated in the aquifer. The temperature anomaly differs from the salinity anomaly by the smaller radius of influence and faster time to reach a new steady-state of the former. Boils discharge water with a temperature equal to the mean groundwater temperature. This influences the yearly and diurnal variation of ditch water temperature in the immediate vicinity of the boil importantly but also the temperature in the downstream direction. Temporary nature of the boil (e.g. stability of the conduit, discharge rate), uncertainty on the 3D construction of the conduit and heterogeneity of the subsoil make it unlikely that temperature measurements can be interpreted further than a qualitative level.

Pretest thermal analysis of the Tuff Water Migration/In-Situ Heater Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bulmer, B.M.

This report describes the pretest thermal analysis for the Tuff Water Migration/In-Situ Heater Experiment to be conducted in welded tuff in G-tunnel, Nevada Test Site. The parametric thermal modeling considers variable boiling temperature, tuff thermal conductivity, tuff emissivity, and heater operating power. For nominal tuff properties, some near field boiling is predicted for realistic operating power. However, the extent of boiling will be strongly determined by the ambient (100% water saturated) rock thermal conductivity. In addition, the thermal response of the heater and of the tuff within the dry-out zone (i.e., bounded by boiling isotherm) is dependent on the temperaturemore » variation of rock conductivity as well as the extent of induced boiling.« less

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, C. S.; Zhang, Hongbin

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

NASA Astrophysics Data System (ADS)

Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

2016-02-01

This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS

DOE PAGES

Brown, C. S.; Zhang, Hongbin

2016-05-24

Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surfacemore » temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.« less

Iron release from the Lucky Iron Fish®: safety considerations.

PubMed

Armstrong, Gavin R; Dewey, Cate E; Summerlee, Alastair Js

2017-01-01

The principal objective was to explore in greater detail safety issues with regard to the use of the Lucky Iron Fish® (fish) as a treatment for iron deficiency and iron deficiency anaemia in women in rural Cambodia. Experiments were done to determine: (1) purity of the iron in the fish by mass spectroscopy; (2) release of iron and contaminants released during boiling in water using inductive-ly-coupled plasma optical emission spectroscopy; (3) the impact of cooking time, acidity and number of fish in acidified water and two types of Khmer soups; and (4) drinkability of the water after boiling with different num-bers of fish. The fish is composed primarily of ferrous iron with less than 12% non-ferrous iron. Contaminants were either not detectable or levels were below the acceptable standards set by the World Health Organization. The length of time boiling the fish and the acidity of the water increased iron release but even with 5 fish boiled for 60 minutes, iron levels only approached levels where side effects are observed. Boiling one fish in water did not affect the perception of colour, smell or taste of the water but boiling in water with two or more fish resulted in the water being unpalatable which further limits the potential for iron toxicity from using the fish. The results suggest that the Lucky Iron Fish™ may be a safe treatment for iron deficiency.

MOX fuel assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reese, A.P.; Crowther, R.L. Jr.

1992-02-18

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons formore » shaping the cold reactivity shutdown zone in the fuel bundle.« less

Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-01-01

This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

PANDA asymmetric-configuration passive decay heat removal test results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fischer, O.; Dreier, J.; Aubert, C.

1997-12-01

PANDA is a large-scale, low-pressure test facility for investigating passive decay heat removal systems for the next generation of LWRs. In the first series of experiments, PANDA was used to examine the long-term LOCA response of the Passive Containment Cooling System (PCCS) for the General Electric (GE) Simplified Boiling Water Reactor (SBWR). The test objectives include concept demonstration and extension of the database available for qualification of containment codes. Also included is the study of the effects of nonuniform distributions of steam and noncondensable gases in the Dry-well (DW) and in the Suppression Chamber (SC). 3 refs., 9 figs.

The Japanese utilities` expectations for subchannel analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toba, Akio; Omoto, Akira

1995-12-01

Boiling water reactor (BWR) utilities in Japan began to consider the development of a mechanistic model to describe the critical heat transfer conditions in the BWR fuel subchannel. Such a mechanistic model will not only decrease the necessity of tests, but will also help by removing some overly conservative safety margins in thermal hydraulics. With the use of a postdryout heat transfer correlation, new acceptance criteria may be applicable to evaluate the fuel integrity. Mechanistic subchannel analysis models will certainly back up this approach. This model will also be applicable to the analysis of large-size fuel bundles and examination ofmore » corrosion behavior.« less

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robb, Kevin R.

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dallman, R J; Gottula, R C; Holcomb, E E

1987-05-01

An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented.

The Hatch-Smolensk exchange

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sproles, A.

1993-03-01

During summer 1992, the World Association of Nuclear Operators (WANO) sponsored an exchange visit between Georgia Power Company's Edwin I. Hatch nuclear plant, a two-unit boiling water reactor site, and the Smolensk atomic energy station, a three-unit RBMK (graphite-moderated and light-water-cooled) plant located 350 km west of Moscow, in Desnogorsk, Russia. The Plant Hatch team included Glenn Goode, manager of engineering support; Curtis Coggin, manager of training and emergency preparedness; Wayne Kirkley, manager of health physics and chemistry; John Lewis, manager of operations; Ray Baker, coordinator of nuclear fuels and contracts; and Bruce McLeod, manager of nuclear maintenance support. Alsomore » traveling with the team was Jerald Towgood, of WANO's Atlanta Centre. The Hatch team visited the Smolensk plant during the week of July 27, 1992.« less

The purity of water at hospital and at home as a problem of intercultural understanding.

PubMed

Burghart, R

1996-03-01

Women in a provincial town in southern Nepal were instructed by medical doctors and compounders to boil water, and to keep it boiling for 15 minutes before mixing it with infant formula or oral rehydration salts. Most women ignored the advice. Those who seemed to follow it merely brought the water to boil. This report describes how and why women boil water and assesses the health implications of their practices. The failure of women to adopt "proper" procedures procedures provides a point of entry into an analysis of the role of intercultural dialogue in exposing one's presuppositions about health and empowering one to change them.

Post-test analysis of dryout test 7B' of the W-1 Sodium Loop Safety Facility Experiment with the SABRE-2P code. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rose, S.D.; Dearing, J.F.

An understanding of conditions that may cause sodium boiling and boiling propagation that may lead to dryout and fuel failure is crucial in liquid-metal fast-breeder reactor safety. In this study, the SABRE-2P subchannel analysis code has been used to analyze the ultimate transient of the in-core W-1 Sodium Loop Safety Facility experiment. This code has a 3-D simple nondynamic boiling model which is able to predict the flow instability which caused dryout. In other analyses dryout has been predicted for out-of-core test bundles and so this study provides additional confirmation of the model.

On the pulse boiling frequency in thermosyphons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, J.F.; Wang, J.C.Y.

1992-02-01

The unsteady periodic boiling phenomenon, pulse boiling, appearing in the evaporator of thermosyphons has been mentioned and investigated by many researchers. The heat transfer coefficient in evaporators was predicted according to different considerations of flow patterns. For instance, Shiraishi et al. proposed a method based on a combination flow pattern: the nucleate boiling in a liquid pool and the evaporation from a falling condensate film. Liu et al. only considered a pure pulse boiling flow pattern, and Xin et al. focused on the flow pattern of the continuous boiling process without pulse phenomenon. Besides, the forming conditions of pulse boilingmore » were also described differently. Xin et al. also reported that pulse boiling cannot occur in a carbon-steel/water heat pipe; Ma et al., however, observed this phenomenon in a carbon-steel/water thermosyphon. Nearly all researchers mentioned that this phenomenon indeed exists in glass/water thermosyphons. Although the influential factors have been discussed qualitatively, the quantitative analysis has yet to be conducted. This study focuses on the pulse boiling frequency as a criterion for the determination of flow patterns, and attempts are made to predict the frequency both experimentally and theoretically.« less

16 CFR 1511.5 - Structural integrity tests.

Code of Federal Regulations, 2012 CFR

2012-01-01

... pounds for an additional 10 seconds. (c) Heat cycle deterioration. After the testing prescribed in... pacifier in boiling water for 5 minutes and then remove the pacifier and allow it to cool for 5 minutes in... in the boiling water for 5 minutes. The process shall be repeated for a total of 6 boiling/cooling...

16 CFR 1511.5 - Structural integrity tests.

Code of Federal Regulations, 2011 CFR

2011-01-01

... pounds for an additional 10 seconds. (c) Heat cycle deterioration. After the testing prescribed in... pacifier in boiling water for 5 minutes and then remove the pacifier and allow it to cool for 5 minutes in... in the boiling water for 5 minutes. The process shall be repeated for a total of 6 boiling/cooling...

16 CFR § 1511.5 - Structural integrity tests.

Code of Federal Regulations, 2013 CFR

2013-01-01

... pounds for an additional 10 seconds. (c) Heat cycle deterioration. After the testing prescribed in... pacifier in boiling water for 5 minutes and then remove the pacifier and allow it to cool for 5 minutes in... in the boiling water for 5 minutes. The process shall be repeated for a total of 6 boiling/cooling...

16 CFR 1511.5 - Structural integrity tests.

Code of Federal Regulations, 2014 CFR

2014-01-01

... pounds for an additional 10 seconds. (c) Heat cycle deterioration. After the testing prescribed in... pacifier in boiling water for 5 minutes and then remove the pacifier and allow it to cool for 5 minutes in... in the boiling water for 5 minutes. The process shall be repeated for a total of 6 boiling/cooling...

Generation of shockwave and vortex structures at the outflow of a boiling water jet

NASA Astrophysics Data System (ADS)

Alekseev, M. V.; Lezhnin, S. I.; Pribaturin, N. A.; Sorokin, A. L.

2014-12-01

Results of numerical simulation for shock waves and generation of vortex structures during unsteady outflow of boiling liquid jet are presented. The features of evolution of shock waves and vortex structures formation during unsteady outflow of boiling water are compared with corresponding structures during unsteady gas outflow.

Microbiological Evaluation of Household Drinking Water Treatment in Rural China Shows Benefits of Electric Kettles: A Cross-Sectional Study

PubMed Central

Cohen, Alasdair; Tao, Yong; Luo, Qing; Zhong, Gemei; Romm, Jeff; Colford, John M.; Ray, Isha

2015-01-01

Background In rural China ~607 million people drink boiled water, yet little is known about prevailing household water treatment (HWT) methods or their effectiveness. Boiling, the most common HWT method globally, is microbiologically effective, but household air pollution (HAP) from burning solid fuels causes cardiovascular and respiratory disease, and black carbon emissions exacerbate climate change. Boiled water is also easily re-contaminated. Our study was designed to identify the HWT methods used in rural China and to evaluate their effectiveness. Methods We used a geographically stratified cross-sectional design in rural Guangxi Province to collect survey data from 450 households in the summer of 2013. Household drinking water samples were collected and assayed for Thermotolerant Coliforms (TTC), and physicochemical analyses were conducted for village drinking water sources. In the winter of 2013–2104, we surveyed 120 additional households and used remote sensors to corroborate self-reported boiling data. Findings Our HWT prevalence estimates were: 27.1% boiling with electric kettles, 20.3% boiling with pots, 34.4% purchasing bottled water, and 18.2% drinking untreated water (for these analyses we treated bottled water as a HWT method). Households using electric kettles had the lowest concentrations of TTC (73% lower than households drinking untreated water). Multilevel mixed-effects regression analyses showed that electric kettles were associated with the largest Log10TTC reduction (-0.60, p<0.001), followed by bottled water (-0.45, p<0.001) and pots (-0.44, p<0.01). Compared to households drinking untreated water, electric kettle users also had the lowest risk of having TTC detected in their drinking water (risk ratio, RR = 0.49, 0.34–0.70, p<0.001), followed by bottled water users (RR = 0.70, 0.53–0.93, p<0.05) and households boiling with pots (RR = 0.74, 0.54–1.02, p = 0.06). Conclusion As far as we are aware, this is the first HWT-focused study in China, and the first to quantify the comparative advantage of boiling with electric kettles over pots. Our results suggest that electric kettles could be used to rapidly expand safe drinking water access and reduce HAP exposure in rural China. PMID:26421716

Microbiological Evaluation of Household Drinking Water Treatment in Rural China Shows Benefits of Electric Kettles: A Cross-Sectional Study.

PubMed

Cohen, Alasdair; Tao, Yong; Luo, Qing; Zhong, Gemei; Romm, Jeff; Colford, John M; Ray, Isha

2015-01-01

In rural China ~607 million people drink boiled water, yet little is known about prevailing household water treatment (HWT) methods or their effectiveness. Boiling, the most common HWT method globally, is microbiologically effective, but household air pollution (HAP) from burning solid fuels causes cardiovascular and respiratory disease, and black carbon emissions exacerbate climate change. Boiled water is also easily re-contaminated. Our study was designed to identify the HWT methods used in rural China and to evaluate their effectiveness. We used a geographically stratified cross-sectional design in rural Guangxi Province to collect survey data from 450 households in the summer of 2013. Household drinking water samples were collected and assayed for Thermotolerant Coliforms (TTC), and physicochemical analyses were conducted for village drinking water sources. In the winter of 2013-2104, we surveyed 120 additional households and used remote sensors to corroborate self-reported boiling data. Our HWT prevalence estimates were: 27.1% boiling with electric kettles, 20.3% boiling with pots, 34.4% purchasing bottled water, and 18.2% drinking untreated water (for these analyses we treated bottled water as a HWT method). Households using electric kettles had the lowest concentrations of TTC (73% lower than households drinking untreated water). Multilevel mixed-effects regression analyses showed that electric kettles were associated with the largest Log10TTC reduction (-0.60, p<0.001), followed by bottled water (-0.45, p<0.001) and pots (-0.44, p<0.01). Compared to households drinking untreated water, electric kettle users also had the lowest risk of having TTC detected in their drinking water (risk ratio, RR = 0.49, 0.34-0.70, p<0.001), followed by bottled water users (RR = 0.70, 0.53-0.93, p<0.05) and households boiling with pots (RR = 0.74, 0.54-1.02, p = 0.06). As far as we are aware, this is the first HWT-focused study in China, and the first to quantify the comparative advantage of boiling with electric kettles over pots. Our results suggest that electric kettles could be used to rapidly expand safe drinking water access and reduce HAP exposure in rural China.

Experimental investigation of time and repeated cycles in nucleate pool boiling of alumina/water nanofluid on polished and machined surfaces

NASA Astrophysics Data System (ADS)

Rajabzadeh Dareh, F.; Haghshenasfard, M.; Nasr Esfahany, M.; Salimi Jazi, H.

2018-06-01

Pool boiling heat transfer of pure water and nanofluids on a copper block has been studied experimentally. Nanofluids with various concentrations of 0.0025, 0.005 and 0.01 vol.% are employed and two simple surfaces (polished and machined copper surface) are used as the heating surfaces. The results indicated that the critical heat flux (CHF) in boiling of fluids on the polished surface is 7% higher than CHF on the machined surface. In the case of machined surface, the heat transfer coefficient (HTC) of 0.01 vol.% nanofluid is about 37% higher than HTC of base fluid, while in the polished surface the average HTC of 0.01% nanofluid is about 19% lower than HTC of the pure water. The results also showed that the boiling time and boiling cycles on the polished surface changes the heat transfer performance. By increasing the boiling time from 5 to 10 min, the roughness enhances about 150%, but by increasing the boiling time to 15 min, the roughness enhancement is only 8%.

Study to evaluate the impact of heat treatment on water soluble vitamins in milk.

PubMed

Asadullah; Khair-un-nisa; Tarar, Omer Mukhtar; Ali, Syed Abdul; Jamil, Khalid; Begum, Askari

2010-11-01

To evaluate the effect of domestic boiling practice on the contents of water soluble vitamins of loose milk and quantitative comparison of these vitamins in Ultra High Temperature (UHT) treated packaged milk with that of boiled loose milk. Loose milk samples were collected from various localities of Karachi city (Pakistan). These samples were boiled in simulated household conditions for 5, 10 and 15 minutes. Ultra High Temperature (UHT) treated packaged milk samples of various brands were obtained from the local market. The aliquots were analyzed for water-soluble vitamins using High Performance Liquid Chromatography (HPLC) technique. The mean values and standard deviations for data were computed and compared as well as level of variations were also determined. Conventional boiling caused destruction of water soluble vitamins in milk i.e. vitamin 81 content in fresh milk decreased from 0.037 mg/100 g to 0.027 mg/100 g after 15 min boiling, whereas vitamin B2 from 0.115 to 0.084 mg/100 g, vitamin B3 0.062 to 0.044 mg/100 g, vitamin B6 0.025 to 0.019 mg/100 g and folic acid 3.38 to 2.40 microg/100 g. This accounted for a post-boiling decrease of about 27, 27, 29, 24 and 36% in vitamins B1, B2, B3, B6 and folic acid respectively. The values for vitamins B1, B2, B3, B6 and folic acid determined in boiled milk were significantly lower than UHT treated packaged milk samples by 25.9, 75.0, 54.5, 63.16 and 38.1% respectively. Conventional boiling caused drastic reduction in vitamin levels of loose milk samples. In comparison to this, UHT milk retained high levels of water soluble B-vitamins. Thus it could be envisaged that UHT treated milk provides better water soluble vitamins' nourishment than conventionally boiled milk (JPMA 60:909; 2010).

Development and Assessment of CTF for Pin-resolved BWR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

Interfacial area transport of steam-water two-phase flow in a vertical annulus at elevated pressures

NASA Astrophysics Data System (ADS)

Ozar, Basar

Analysis of accident scenarios in nuclear reactors are done by using codes such as TRACE and RELAP5. Large oscillations in the core void fraction are observed in calculations of advanced passive light water reactors (ALWRs), especially during the low pressure long-term cooling phase. These oscillations are attributed to be numerical in nature and served to limit the accuracy as well as the credibility of the calculations. One of the root causes of these unphysical oscillations is determined to be flow regime transitions caused by the usage of static flow regime maps. The interfacial area transport equation was proposed earlier in order to address these issues. Previous research successfully developed the foundation of the interfacial area transport equation and the experimental techniques needed for the measurement of interfacial area, bubble diameters and velocities. In the past, an extensive database has been then generated for adiabatic air-water conditions in vertical upward and downward bubbly-churn turbulent flows in pipes. Using this database, mechanistic models for the creation (bubble breakup) and destruction (bubble coalescence) of interfacial area have been developed for the bubblyslug flow regime transition. However, none of these studies investigated the effect of phase change. To address this need, a heated annular test section was designed and constructed. The design relied on a three level scaling approach: geometric scaling; hydrodynamic scaling; thermal scaling. The test section consisted of a heated and unheated section in order to study the sub-cooled boiling and bulk condensation/flashing and evaporation phenomena, respectively. Steam-water two-phase flow tests were conducted under sub-cooled boiling conditions in the heated section and with sub-cooled/super-heated bulk liquid in the unheated section. The modeling of interfacial area transport equation with phase change effects was introduced and discussed. Constitutive relations, which took phase change effects into account, for interfacial area transport equation were proposed and implemented. Effects of these constitutive relations on the prediction capability of the transport equation were discussed.

Lattice Boltzmann Methods to Address Fundamental Boiling and Two-Phase Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uddin, Rizwan

2012-01-01

This report presents the progress made during the fourth (no cost extension) year of this three-year grant aimed at the development of a consistent Lattice Boltzmann formulation for boiling and two-phase flows. During the first year, a consistent LBM formulation for the simulation of a two-phase water-steam system was developed. Results of initial model validation in a range of thermo-dynamic conditions typical for Boiling Water Reactors (BWRs) were shown. Progress was made on several fronts during the second year. Most important of these included the simulation of the coalescence of two bubbles including the surface tension effects. Work during themore » third year focused on the development of a new lattice Boltzmann model, called the artificial interface lattice Boltzmann model (AILB model) for the 3 simulation of two-phase dynamics. The model is based on the principle of free energy minimization and invokes the Gibbs-Duhem equation in the formulation of non-ideal forcing function. This was reported in detail in the last progress report. Part of the efforts during the last (no-cost extension) year were focused on developing a parallel capability for the 2D as well as for the 3D codes developed in this project. This will be reported in the final report. Here we report the work carried out on testing the AILB model for conditions including the thermal effects. A simplified thermal LB model, based on the thermal energy distribution approach, was developed. The simplifications are made after neglecting the viscous heat dissipation and the work done by pressure in the original thermal energy distribution model. Details of the model are presented here, followed by a discussion of the boundary conditions, and then results for some two-phase thermal problems.« less

Microbiological effectiveness and cost of disinfecting water by boiling in semi-urban India.

PubMed

Clasen, Thomas; McLaughlin, Catherine; Nayaar, Neeru; Boisson, Sophie; Gupta, Romesh; Desai, Dolly; Shah, Nimish

2008-09-01

Despite shortcomings, boiling is the most common means of treating water at home and the benchmark against which emerging point-of-use water treatment approaches are measured. In a 5-month study, we assessed the microbiological effectiveness and cost of the practice among 218 self-reported boilers relying on unprotected water supplies. Boiling was associated with a 99% reduction in geometric mean fecal coliforms (FCs; P < 0.001). Despite high levels of fecal contamination in source water, 59.6% of stored drinking water samples from self-reported boilers met the World Health Organization standard for safe drinking water (0 FC/100 mL), and 5.7% were between 1 and 10 FC/100 mL. Nevertheless, 40.4% of stored drinking water samples were positive for FCs, with 25.1% exceeding 100 FC/100 mL. The estimated monthly fuel cost for boiling was INR 43.8 (US$0.88) for households using liquid petroleum gas and INR 34.7 (US$0.69) for households using wood.

Comparison of boiling and chlorination on the quality of stored drinking water and childhood diarrhoea in Indonesian households

PubMed Central

FAGERLI, K.; TRIVEDI, K. K.; SODHA, S. V.; BLANTON, E.; ATI, A.; NGUYEN, T.; DELEA, K. C.; AINSLIE, R.; FIGUEROA, M. E.; KIM, S.; QUICK, R.

2018-01-01

SUMMARY We compared the impact of a commercial chlorination product (brand name Air RahMat) in stored drinking water to traditional boiling practices in Indonesia. We conducted a baseline survey of all households with children <5 years in four communities, made 11 subsequent weekly home visits to assess acceptability and use of water treatment methods, measured Escherichia coli concentration in stored water, and determined diarrhoea prevalence among children <5 years. Of 281 households surveyed, boiling (83%) and Air RahMat (7%) were the principal water treatment methods. Multivariable log-binomial regression analyses showed lower risk of E. coli in stored water treated with Air RahMat than boiling (risk ratio (RR) 0·75, 95% confidence interval (CI) 0·56–1·00). The risk of diarrhoea in children <5 years was lower among households using Air RahMat (RR 0·43, 95% CI 0·19–0·97) than boiling, and higher in households with E. coli concentrations of 1–1000 MPN/100 ml (RR 1·54, 95% CI 1·04–2·28) or >1000 MPN/100 ml (RR 1·86, 95% CI 1·09–3·19) in stored water than in households without detectable E. coli. Although results suggested that Air RahMat water treatment was associated with lower E. coli contamination and diarrhoeal rates among children <5 years than water treatment by boiling, Air RahMat use remained low. PMID:28942755

A Novel Role of Three Dimensional Graphene Foam to Prevent Heater Failure during Boiling

PubMed Central

Ahn, Ho Seon; Kim, Ji Min; Park, Chibeom; Jang, Ji-Wook; Lee, Jae Sung; Kim, Hyungdae; Kaviany, Massoud; Kim, Moo Hwan

2013-01-01

We report a novel boiling heat transfer (NBHT) in reduced graphene oxide (RGO) suspended in water (RGO colloid) near critical heat flux (CHF), which is traditionally the dangerous limitation of nucleate boiling heat transfer because of heater failure. When the heat flux reaches the maximum value (CHF) in RGO colloid pool boiling, the wall temperature increases gradually and slowly with an almost constant heat flux, contrary to the rapid wall temperature increase found during water pool boiling. The gained time by NBHT would provide the safer margin of the heat transfer and the amazing impact on the thermal system as the first report of graphene application. In addition, the CHF and boiling heat transfer performance also increase. This novel boiling phenomenon can effectively prevent heater failure because of the role played by the self-assembled three-dimensional foam-like graphene network (SFG). PMID:23743619

Physical quality of Simental Ongole crossbred silverside meat at various boiling times

NASA Astrophysics Data System (ADS)

Riyanto, J.; Cahyadi, M.; Guntari, W. S.

2018-03-01

This study aims to determine the physical quality of silverside beef meat at various boiling times. Samples that have been used are the back thigh or silverside meat. Treatment of boiling meat included TR (meat without boiled), R15 (boiled 15 minutes), and R30 (boiled for 30 minutes). The experimental design using Completely Randomized Design with 3 replications. Each replication was done in triple physical quality test. Determination of physical quality was performed at the Livestock Industry and Processing Laboratory at Sebelas Maret University Surakarta and the Meat Technology Laboratory at the Faculty of Animal Husbandry of Gadjah Mada University. The result of variance analysis showed that boiling affect cooking loss (P≥0.05) and but did not affect (P≤0,05) pH, water holding capacity and meat tenderness. The conclusions of the study showed that boiling for 15 minutes and 30 minutes decreased the cooking loss of Simental Ongole Crossbred silverside meat. Meat physical quality of pH, water holding capacity and the value of tenderness is not affected by boiling for 15 and 30 minutes.

The Amateur Scientist.

ERIC Educational Resources Information Center

Walker, Jearl

1982-01-01

Discusses the thermodynamics of heating and boiling water, including the mechanics of bubble formation. Indicates that when salt was added to boiling water (when cooking pasta) the water temperature rose less than a degree, concluding that salt serves only as a seasoning. (Author/JN)

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Overnight soaking or boiling of "Matooke" to reduce potassium content for patients with chronic kidney disease: does it really work?

PubMed

Asiimwe, J; Sembajwe, L F; Senoga, A; Bakiika, E; Muwonge, H; Kalyesubula, R

2013-09-01

There is an increase in number of patients with chronic kidney disease (CKD) in Uganda's health facilities looking for different options of preparing matooke (bananas), their staple food. To establish and evaluate an effective method of removing potassium from bananas (matooke). Bananas were sampled from 5 markets in Kampala, Uganda. Deionized water was used to soak the bananas and the potassium concentration was determined using an atomic absorption spectrophotometer in both the bananas and water after soaking for varying time intervals. We also determined the potassium concentrations in the bananas and the water after boiling the bananas at 200 degrees Celsius at intervals of 10 minutes (for 60 minutes). The potassium concentration did not appear to change on soaking alone without boiling. However, on boiling, the concentration in the bananas decreased from about 1.4 ppm to approx. 1 ppm after 60 min; yet the concentration of potassium released into deionized water increased steadily from 0.0 ppm to about 1.2 ppm after 60 min of boiling. This study demonstrates that boiling the bananas is a more effective way of removing the potassium from bananas than simply soaking them.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, Norman B.

1998-01-01

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

Nuclear fuel elements made from nanophase materials

DOEpatents

Heubeck, N.B.

1998-09-08

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Pool boiling of water-Al2O3 and water-Cu nanofluids on horizontal smooth tubes

PubMed Central

2011-01-01

Experimental investigation of heat transfer during pool boiling of two nanofluids, i.e., water-Al2O3 and water-Cu has been carried out. Nanoparticles were tested at the concentration of 0.01%, 0.1%, and 1% by weight. The horizontal smooth copper and stainless steel tubes having 10 mm OD and 0.6 mm wall thickness formed test heater. The experiments have been performed to establish the influence of nanofluids concentration as well as tube surface material on heat transfer characteristics at atmospheric pressure. The results indicate that independent of concentration nanoparticle material (Al2O3 and Cu) has almost no influence on heat transfer coefficient while boiling of water-Al2O3 or water-Cu nanofluids on smooth copper tube. It seems that heater material did not affect the boiling heat transfer in 0.1 wt.% water-Cu nanofluid, nevertheless independent of concentration, distinctly higher heat transfer coefficient was recorded for stainless steel tube than for copper tube for the same heat flux density. PMID:21711741

Estimating surface temperature in forced convection nucleate boiling - A simplified method

NASA Technical Reports Server (NTRS)

Hendricks, R. C.; Papell, S. S.

1977-01-01

A simplified expression to estimate surface temperatures in forced convection boiling was developed using a liquid nitrogen data base. Using the principal of corresponding states and the Kutateladze relation for maximum pool boiling heat flux, the expression was normalized for use with other fluids. The expression was applied also to neon and water. For the neon data base, the agreement was acceptable with the exclusion of one set suspected to be in the transition boiling regime. For the water data base at reduced pressure greater than 0.05 the agreement is generally good. At lower reduced pressures, the water data scatter and the calculated temperature becomes a function of flow rate.

Two-phase pressure drop reduction BWR assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dix, G.E.; Crowther, R.L.; Colby, M.J.

1992-05-12

This patent describes a boiling water reactor having discrete bundles of fuel rods confined within channel enclosed fuel assemblies, an improvement to a fuel bundle assembly for placement in the reactor. It comprises a fuel channel having vertically extending walls forming a continuous channel around a fuel assembly volume, the channel being open at the bottom end for engagement to a lower tie plate and open at the upper end for engagement to an upper tie plate; rods for placement within the chamber, each the rod containing fissile material for producing nuclear reaction when in the presence of sufficient moderatedmore » neutron flux; a lower tie plate for supporting the bundle of rods within the channel, the lower tie plate for supporting the bundle of rods within the channel, the lower tie plate joining the bottom of the channel to close the bottom end of the channel, the lower tie plate providing defined apertures for the inflow of water in the channel between the rods for the generating of steam during the nuclear reaction; the plurality of fuel rods extending from the lower tie plate wherein a single phase region of the water in the bundle is defined to an upward portion of the bundle wherein a two phase region of the water and steam in the bundle is defined during nuclear steam generating reaction in the fuel bundle.« less

Energy-efficient drinking water disinfection for greenhouse gas mitigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gadgil, A.J.; Greene, D.M.; Rosenfeld, A.

Anecdotal evidence suggests that approximately one billion people worldwide use cookstoves to boil their drinking water. About half of this population is in China. Some populations (e.g. Jakarta) spend 1% of their GDP on boiling drinking water. Impoverished and/or ignorant populations not yet boiling their drinking water will do so when they can both afford it and understand the risks of unsafe drinking water. A recently developed water disinfection technology (UV Waterworks) can produce safe drinking water while earning tradable carbon credits (or credit as a clean development mechanism) when implemented as part of national energy, health, and carbon emissionsmore » trading policy, UV Waterworks uses approximately 6,000 times less energy than boiling over a biomass cookstove. Each unit that replaces boiling may save up to 175 or 300 tons/year of carbon-equivalent GHG emissions, depending on if it replaces sustainably harvested biomass (SHB) or non-SHB. For the approximately 500M Chinese boiling their drinking water over biomass (assumed SHB), this suggests a technical potential (that is, potential under the limiting case of 100% market adoption) of saving 87M tons/year of carbon-equivalent non-CO{sub 2} GHG emissions. The energy savings and corresponding emissions reductions will vary with cookstove fuels and stove efficiency: non-SHB and kerosene represent the most and least GHG-producing cookstove fuels, respectively, among those readily available to the populations of interest. The authors bracket the global technical potential for carbon emission reductions resulting from implementation of UV Waterworks, and estimate the value of tradable carbon credits earned from these reductions.« less

Development of new UV-I. I. Cerenkov Viewing Device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki; Nemoto, Koshichi

1994-02-01

The Cerenkov glow images from boiling-water reactors (BWR) and pressurized-water reactors (PWR) irradiated fuel assemblies are generally used for inspections. However, sometimes it is difficult or impossible to identify the image by the conventional Cerenkov Viewing Device (CVD), because of the long cooling time and/or low burnup. Now a new UV-I.I. (Ultra-Violet light Image Intensifier) CVD has been developed, which can detect the very weak Cerenkov glow from spent fuel assemblies. As this new device uses the newly developed proximity focused type UV-I.I., Cerenkov photons are used efficiently, producing better quality Cerenkov glow images. Moreover, since the image is convertedmore » to a video signal, it is easy to improve the signal to noise ratio (S/N) by an image processor. The new CVD was tested at BWR and PWR power plants in Japan, with fuel burnups ranging from 6,200--33,000 MWD/MTU (megawatt days per metric ton of uranium) and cooling times ranging from 370 to 6,200 d. The tests showed that the new CVD is superior to the conventional STA/CRIEPI CVD, and could detect very feeble Cerenkov glow images using an image processor.« less

Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ernst, Frank

We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, andmore » fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.« less

Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are themore » MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.« less

Decontamination and decommissioning of the BORAX-V leach pond. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, D.L.

1985-01-01

This report describes the decontamination and decommissioning (D and D) of the BORAX-V leach pond located at the Idaho National Engineering Laboratory (INEL). The leach pond became radioactively contaminated from the periodic discharge of low-level liquid waste during operation of the Boiling Water Reactor Experiments (BORAX) from 1954 to 1964. This report describes work performed to accomplish the D and D objectives of stabilizing the leach pond and preventing the spread of contamination. D and D of the BORAX-V leach pond consisted to backfilling the pond with clean soil, grading and seeding the area, and erecting a permanent marker tomore » identify very low-level subsurface contamination.« less

Benchmark calculation for radioactivity inventory using MAXS library based on JENDL-4.0 and JEFF-3.0/A for decommissioning BWR plants

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi

2016-06-01

We performed benchmark calculation for radioactivity activated in a Primary Containment Vessel (PCV) of a Boiling Water Reactor (BWR) by using MAXS library, which was developed by collapsing with neutron energy spectra in the PCV of the BWR. Radioactivities due to neutron irradiation were measured by using activation foil detector of Gold (Au) and Nickel (Ni) at thirty locations in the PCV. We performed activation calculations of the foils with SCALE5.1/ORIGEN-S code with irradiation conditions of each foil location as the benchmark calculation. We compared calculations and measurements to estimate an effectiveness of MAXS library.

System Study: Isolation Condenser 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the isolation condenser (ISO) system at four U.S. boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trends were identified. A statistically significant decreasing trend was identified for ISO unreliability. The magnitude of the trend indicated a 1.5 percent decrease inmore » system unreliability over the last 10 years.« less

Comparison of boiling and chlorination on the quality of stored drinking water and childhood diarrhoea in Indonesian households.

PubMed

Fagerli, K; Trivedi, K K; Sodha, S V; Blanton, E; Ati, A; Nguyen, T; Delea, K C; Ainslie, R; Figueroa, M E; Kim, S; Quick, R

2017-11-01

We compared the impact of a commercial chlorination product (brand name Air RahMat) in stored drinking water to traditional boiling practices in Indonesia. We conducted a baseline survey of all households with children 1000 MPN/100 ml (RR 1·86, 95% CI 1·09-3·19) in stored water than in households without detectable E. coli. Although results suggested that Air RahMat water treatment was associated with lower E. coli contamination and diarrhoeal rates among children <5 years than water treatment by boiling, Air RahMat use remained low.

Food-cooking processes modulate allergenic properties of hen's egg white proteins.

PubMed

Liu, Xiaoyu; Feng, Bai-Sui; Kong, Xiaoli; Xu, Hong; Li, Xiumin; Yang, Ping-Chang; Liu, Zhigang

2013-01-01

Reducing the allergenicity of food allergens can suppress the clinical symptoms of food allergy. The objective of the present study was to investigate the effects of processing on the allergenic properties of hen's egg white proteins. Eggs were processed by traditional Chinese cooking, including steaming, water boiling, frying, spicing and tea boiling. The contents of processed egg protein were assessed by sodium dodecyl sulfate polyacrylamide gel electrophoresis; the allergenicity was evaluated by Western blotting, enzyme-linked immunosorbent assay and enzyme allergosorbent test inhibition. Circular dichroism spectrum analysis of four major egg allergens from various egg products was performed as well. A mouse model of food allergy was developed to test the allergenicity of processed egg protein in vivo. Protein degradation was significant following tea boiling and spiced-tea boiling. The total allergenic potential of water-boiled egg and fried egg was relatively higher than that of steamed egg, spiced egg and tea-boiled egg. Challenge with proteins from raw egg, water-boiled egg and fried egg induced skewed T-helper 2 pattern responses (Th2 responses) in the intestine of mice sensitized to egg proteins; however, when the mice sensitized to egg proteins were challenged with proteins from steamed egg, spiced egg and tea-boiled egg, respectively, only weak Th2 responses were induced in their intestine. Processing by steaming, spicing, or tea boiling can weaken the allergenicity of egg proteins. Copyright © 2012 S. Karger AG, Basel.

Hydrolysis of Glycosidic Flavonoids during the Preparation of Danggui Buxue Tang: An Outcome of Moderate Boiling of Chinese Herbal Mixture

PubMed Central

Zhang, Wendy Li; Chen, Jian-Ping; Lam, Kelly Yin-Ching; Zhan, Janis Ya-Xian; Yao, Ping; Dong, Tina Ting-Xia; Tsim, Karl Wah-Keung

2014-01-01

Chemical change during boiling of herbal mixture is a puzzle. By using Danggui Buxue Tang (DBT), a herbal decoction that contains Astragali Radix (AR) and Angelicae Sinensis Radix (ASR), we developed a model in analyzing the hydrolysis of flavonoid glycosides during the boiling of herbal mixture in water. A proper preparation of DBT is of great benefit to the complete extraction of bioactive ingredients. Boiling of DBT in water increased the solubility of AR-derived astragaloside IV, calycosin, formononetin, calycosin-7-O-β-D-glucoside, and ononin in a time- and temperature-dependent manner: the amounts of these chemicals reached a peak at 2 h. The glycosidic resides of AR, calycosin-7-O-β-D-glucoside, and ononin could be hydrolyzed during the moderate boiling process to form calycosin and formononetin, respectively. The hydrolysis efficiency was strongly affected by pH, temperature, and amount of herbs. Interestingly, the preheated herbs were not able to show this hydrolytic activity. The current results supported the rationality of ancient preparation of DBT in boiling water by moderate heat. PMID:24744813

Pool boiling of nanofluids on rough and porous coated tubes: experimental and correlation

NASA Astrophysics Data System (ADS)

Cieśliński, Janusz T.; Kaczmarczyk, Tomasz Z.

2014-06-01

The paper deals with pool boiling of water-Al2O3 and water- Cu nanofluids on rough and porous coated horizontal tubes. Commercially available stainless steel tubes having 10 mm outside diameter and 0.6 mm wall thickness were used to fabricate the test heater. The tube surface was roughed with emery paper 360 or polished with abrasive compound. Aluminium porous coatings of 0.15 mm thick with porosity of about 40% were produced by plasma spraying. The experiments were conducted under different absolute operating pressures, i.e., 200, 100, and 10 kPa. Nanoparticles were tested at the concentration of 0.01, 0.1, and 1% by weight. Ultrasonic vibration was used in order to stabilize the dispersion of the nanoparticles. It was observed that independent of operating pressure and roughness of the stainless steel tubes addition of even small amount of nanoparticles augments heat transfer in comparison to boiling of distilled water. Contrary to rough tubes boiling heat transfer coefficient of tested nanofluids on porous coated tubes was lower compared to that for distilled water while boiling on porous coated tubes. A correlation equation for prediction of the average heat transfer coefficient during boiling of nanofluids on smooth, rough and porous coated tubes is proposed. The correlation includes all tested variables in dimensionless form and is valid for low heat flux, i.e., below 100 kW/m2.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

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 amore » 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.« less

75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-13

...The Food and Drug Administration (FDA) is announcing the availability of a guidance entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory.'' This guidance is intended to advise food manufacturers that once a boil-water advisory has been issued they should stop using the water subject to the advisory until the water again meets the applicable Federal and State drinking water quality standards. Further, this guidance is intended to assist food manufacturers in evaluating food that already was produced with water subject to the advisory. The guidance is in response to the recent major water pipe break in Massachusetts that interrupted service to 30 Massachusetts Water Resources Authority (MWRA) customer communities (serving approximately 2 million residents).

Propagation of neutron-reaction uncertainties through multi-physics models of novel LWR's

NASA Astrophysics Data System (ADS)

Hernandez-Solis, Augusto; Sjöstrand, Henrik; Helgesson, Petter

2017-09-01

The novel design of the renewable boiling water reactor (RBWR) allows a breeding ratio greater than unity and thus, it aims at providing for a self-sustained fuel cycle. The neutron reactions that compose the different microscopic cross-sections and angular distributions are uncertain, so when they are employed in the determination of the spatial distribution of the neutron flux in a nuclear reactor, a methodology should be employed to account for these associated uncertainties. In this work, the Total Monte Carlo (TMC) method is used to propagate the different neutron-reactions (as well as angular distributions) covariances that are part of the TENDL-2014 nuclear data (ND) library. The main objective is to propagate them through coupled neutronic and thermal-hydraulic models in order to assess the uncertainty of important safety parameters related to multi-physics, such as peak cladding temperature along the axial direction of an RBWR fuel assembly. The objective of this study is to quantify the impact that ND covariances of important nuclides such as U-235, U-238, Pu-239 and the thermal scattering of hydrogen in H2O have in the deterministic safety analysis of novel nuclear reactors designs.

Self-Sustaining Thorium Boiling Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare themore » RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.« less

Waste Package Component Design Methodology Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.C. Mecham

2004-07-12

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and usemore » of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.« less

Quenching of a highly superheated porous medium by injection of water

NASA Astrophysics Data System (ADS)

Fichot, F.; Bachrata, A.; Repetto, G.; Fleurot, J.; Quintard, M.

2012-11-01

Understanding of two-phase flow through porous medium with intense phase change is of interest in many situations, including nuclear, chemical or geophysical applications. Intense boiling occurs when the liquid is injected into a highly superheated medium. Under such conditions, the heat flux extracted by the fluid from the porous medium is mainly governed by the nucleation of bubbles and by the evaporation of thin liquid films. Both configurations are possible, depending on local flow conditions and on the ratio of bubble size to pore size. The present study is motivated by the safety evaluation of light water nuclear reactors in case of a severe accident scenario, such as the one that happened in Fukushima Dai-ichi plant in March, 2011. If water sources are not available for a long period of time, the reactor core heats up due to the residual power and eventually becomes significantly damaged due to intense oxidation of metals and fragmentation of fuel rods resulting in the formation of a porous medium where the particles have a characteristic length-scale of 1 to 5 mm. The coolability of the porous medium will depend on the water flow rate which can enter the medium under the available driving head and on the geometrical features of the porous matrix (average pore size, porosity). Therefore, it is of high interest to evaluate the conditions for which the injection of water in such porous medium is likely to stop the progression of the accident. The present paper addresses the issue of modelling two-phase flow and heat transfers in a porous medium initially dry, where water is injected. The medium is initially at a temperature well above the saturation temperature of water. In a first part, a summary of existing knowledge is provided, showing the scarcity of models and experimental data. In a second part, new experimental results obtained in an IRSN facility are analysed. The experiment consists in a bed of steel particles that are heated up to 700°Cbefore injecting water. The facility is briefly described. The velocity of the "quench front" (location where particles are quickly cooled down) and the total pressure drop across the medium are estimated. The dependencies of those quantities with respect to the inlet water flow rate, the initial temperature of the medium and the diameter of particles are obtained. In a third part, a model is proposed, based on a previously developed model which is improved in order to take into account intense boiling regimes (in particular nucleate boiling). The model includes a function that takes into account the contact area between water and the particles which depends on the temperature of particles and on the void fraction. That function affects the local intensity of phase change. The model involves a few parameters which cannot be evaluated analytically. Those parameters are bounded, following the analysis of experimental data. Finally, the model is assessed by comparison of calculations with those new experimental data. The satisfactory agreement shows that the model is almost predictive in the range of parameters studied. The experimental results also show that the quench front becomes unstable under certain conditions. This is also analysed and compared with the predictions of the model.

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Philip E. MacDonald

2005-01-01

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission ofmore » the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.« less

Oceanographic effects of the 1992 Point Loma sewage pipe spill

DOE Office of Scientific and Technical Information (OSTI.GOV)

Casey, R.; Ciccateri, A.; Dougherty, K.

Early in early 1992, 180 million gallons of advanced primarily treated sewage emptied into 10 meters of water from the broken Point Loma sewage pipe, San Diego. For about two months a sewage boil about the size of a football field existed at the surface and within the Point Loma kelp bed. Sampling and observations taken during the spill indicated the surface waters at the spill site were grayish and smelling of sewage. The sewage water had mixed with the marine waters reducing salinity to about one-half normal (or 15 ppt.). The sediment load of the sewage coated the bladesmore » of the giant kelp and the kelp was limp and withdrawn from the surface. At the site of the main boil the kelp appeared to have dropped to the bottom. Sediments on the bottom in the boil area were mainly coarse sands as compared to the surrounding sandy-muds. Preliminary results using laboratory analysis suggest: one month into the spill no infauna were observed in the sediments or planktons in the water of the boil area, but were in the surrounding sediments and water; the observed phytoplankton were dominated by dinoflagellates and suggested red tide conditions surrounding the boil. The site has been monitored monthly since the spill to observe further impact and recovery.« less

Infrared thermometry study of nanofluid pool boiling phenomena

PubMed Central

2011-01-01

Infrared thermometry was used to obtain first-of-a-kind, time- and space-resolved data for pool boiling phenomena in water-based nanofluids with diamond and silica nanoparticles at low concentration (<0.1 vol.%). In addition to macroscopic parameters like the average heat transfer coefficient and critical heat flux [CHF] value, more fundamental parameters such as the bubble departure diameter and frequency, growth and wait times, and nucleation site density [NSD] were directly measured for a thin, resistively heated, indium-tin-oxide surface deposited onto a sapphire substrate. Consistent with other nanofluid studies, the nanoparticles caused deterioration in the nucleate boiling heat transfer (by as much as 50%) and an increase in the CHF (by as much as 100%). The bubble departure frequency and NSD were found to be lower in nanofluids compared with water for the same wall superheat. Furthermore, it was found that a porous layer of nanoparticles built up on the heater surface during nucleate boiling, which improved surface wettability compared with the water-boiled surfaces. Using the prevalent nucleate boiling models, it was possible to correlate this improved surface wettability to the experimentally observed reductions in the bubble departure frequency, NSD, and ultimately to the deterioration in the nucleate boiling heat transfer and the CHF enhancement. PMID:21711754

NUCLEAR POWER PLANT

DOEpatents

Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

1963-05-14

A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

Decommissioning of German Nuclear Research Facilities under the Governance of the Federal Ministry of Education and Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weigl, M.

2008-07-01

Since the announcement of the first nuclear program in 1956, nuclear R and D in Germany has been supported by the Federal Government under four nuclear programs and later on under more general energy R and D programs. The original goal was to help German industry to achieve safe, low-cost generation of energy and self-sufficiency in the various branches of nuclear technology, including the fast breeder reactor and the fuel cycle. Several national research centers were established to host or operate experimental and demonstration plants. These are mainly located at the sites of the national research centers at Juelich andmore » Karlsruhe. In the meantime, all these facilities were shut down and most of them are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. For two other projects the return to 'green field' sites will be reached by the end of this decade. These are the dismantling of the Multi-Purpose Research Reactor (MZFR) and the Compact Sodium Cooled Reactor (KNK) both located at the Forschungszentrum Karlsruhe. Within these projects a lot of new solutions und innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). For example, high performance underwater cutting technologies like plasma arc cutting and contact arc metal cutting. (authors)« less

Using noble gases measured in spring discharge to trace hydrothermal processes in the Norris Geyser Basin, Yellowstone National Park, U.S.A.

USGS Publications Warehouse

Gardner, W.P.; Susong, D.D.; Solomon, D.K.; Heasler, H.P.

2010-01-01

Dissolved noble gas concentrations in springs are used to investigate boiling of hydrothermal water and mixing of hydrothermal and shallow cool water in the Norris Geyser Basin area. Noble gas concentrations in water are modeled for single stage and continuous steam removal. Limitations on boiling using noble gas concentrations are then used to estimate the isotopic effect of boiling on hydrothermal water, allowing the isotopic composition of the parent hydrothermal water to be determined from that measured in spring. In neutral chloride springs of the Norris Geyser Basin, steam loss since the last addition of noble gas charged water is less than 30% of the total hydrothermal discharge, which results in an isotopic shift due to boiling of ?? 2.5% ??D. Noble gas concentrations in water rapidly and predictably change in dual phase systems, making them invaluable tracers of gas-liquid interaction in hydrothermal systems. By combining traditional tracers of hydrothermal flow such as deuterium with dissolved noble gas measurements, more complex hydrothermal processes can be interpreted. ?? 2010 Elsevier B.V.

Production induced boiling and cold water entry in the Cerro Prieto geothermal reservoir indicated by chemical and physical measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grant, M.A.; Truesdell, A.H.; Manon, A.

1981-01-01

Chemical and physical data suggest that the relatively shallow western part of the Cerro Prieto reservoir is bounded below by low permeability rocks, and above and at the sides by an interface with cooler water. There is no continuous permeability barrier around or immediately above the reservoir. Permeability within the reservoir is dominantly intergranular. Mixture with cooler water rather than boiling is the dominant cooling process in the natural state, and production causes displacement of hot water by cooler water, not by vapor. Local boiling occurs near most wells in response to pressure decreases, but no general vapor zone hasmore » formed.« less

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Micro-bubble emission boiling with the cavitation bubble blow pit

PubMed Central

Inada, Shigeaki; Shinagawa, Kazuaki; Illias, Suhaimi Bin; Sumiya, Hiroyuki; Jalaludin, Helmisyah A.

2016-01-01

The miniaturization boiling (micro-bubble emission boiling [MEB]) phenomenon, with a high heat removal capacity that contributes considerably to the cooling of the divertor of the nuclear fusion reactor, was discovered in the early 1980s. Extensive research on MEB has been performed since its discovery. However, the progress of the application has been delayed because the generation mechanism of MEB remains unclear. Reasons for this lack of clarity include the complexity of the phenomenon itself and the high-speed phase change phenomenon in which boiling and condensation are rapidly generated. In addition, a more advanced thermal technique is required to realize the MEB phenomenon at the laboratory scale. To the authors’ knowledge, few studies have discussed the rush mechanism of subcooled liquid to the heating surface, which is critical to elucidating the mechanism behind MEB. This study used photographic images to verify that the cavitation phenomenon spreads to the inside of the superheated liquid on the heating surface and thus clarify the mechanism of MEB. PMID:27628271

Science 101: Why Does It Take Longer to Boil Potatoes at High Altitudes?

ERIC Educational Resources Information Center

Robertson, Bill

2017-01-01

Why Does It Take Longer to Boil Potatoes at High Altitudes? This column provides background science information for elementary teachers. This month's issue looks at why water boils at different temperatures at different altitudes.

Effects of Boiling Drinking Water on Diarrhea and Pathogen-Specific Infections in Low- and Middle-Income Countries: A Systematic Review and Meta-Analysis

PubMed Central

Cohen, Alasdair; Colford, John M.

2017-01-01

Abstract. Globally, approximately 2 billion people lack microbiologically safe drinking water. Boiling is the most prevalent household water treatment method, yet evidence of its health impact is limited. To conduct this systematic review, we searched four online databases with no limitations on language or publication date. Studies were eligible if health outcomes were measured for participants who reported consuming boiled and untreated water. We used reported and calculated odds ratios (ORs) and random-effects meta-analysis to estimate pathogen-specific and pooled effects by organism group and nonspecific diarrhea. Heterogeneity and publication bias were assessed using I2, meta-regression, and funnel plots; study quality was also assessed. Of the 1,998 records identified, 27 met inclusion criteria and reported extractable data. We found evidence of a significant protective effect of boiling for Vibrio cholerae infections (OR = 0.31, 95% confidence interval [CI] = 0.13–0.79, N = 4 studies), Blastocystis (OR = 0.35, 95% CI = 0.17–0.69, N = 3), protozoal infections overall (pooled OR = 0.61, 95% CI = 0.43–0.86, N = 11), viral infections overall (pooled OR = 0.83, 95% CI = 0.7–0.98, N = 4), and nonspecific diarrheal outcomes (OR = 0.58, 95% CI = 0.45–0.77, N = 7). We found no evidence of a protective effect for helminthic infections. Although our study was limited by the use of self-reported boiling and non-experimental designs, the evidence suggests that boiling provides measureable health benefits for pathogens whose transmission routes are primarily water based. Consequently, we believe a randomized controlled trial of boiling adherence and health outcomes is needed. PMID:29016318

Heat transfer enhancement at increasing water concentration in alcohol in the process of non-stationary film boiling

NASA Astrophysics Data System (ADS)

Kanin, P. K.; Ryazantsev, V. A.; Lexin, M. A.; Zabirov, A. R.; Yagov, V. V.

2018-03-01

New experimental data on heat transfer in pool film boiling of subcooled ethanol-water mixtures at spherical surfaces are considered. The water solutions with ethanol mass fraction from 10 to 91% and temperature of liquid 50°C were examined. All the experiments were conducted under atmospheric pressure, using the stainless steel sphere of 39 mm in diameter as a cooled body. The sphere was heated up to 450-750°C, depending on ethanol concentration, and immersed into the experimental vessel with subcooled mixture. As it is expected, boiling heat transfer intensifies with ethanol concentration decrease, and duration of cooling decreases. It means that stable film boiling duration decreases, and earlier transition to intensive heat transfer regime occurs.

A geochemical model of the Platanares geothermal system, Honduras

USGS Publications Warehouse

Janik, C.J.; Truesdell, A.H.; Goff, F.; Shevenell, L.; Stallard, M.L.; Trujillo, P.E.; Counce, D.

1991-01-01

Results of exploration drilling combined with results of geologic, geophysical, and hydrogeochemical investigations have been used to construct a geochemical model of the Platanares geothermal system, Honduras. Three coreholes were drilled, two of which produced fluids from fractured Miocene andesite and altered Cretaceous to Eocene conglomerate at 450 to 680 m depth. Large volume artesian flows of 160-165??C, predominantly bicarbonate water are chemically similar to, but slightly less saline than widespread boiling hot-spring waters. The chemistry of the produced fluid is dominated by equilibrium reactions in sedimentary rocks at greater depths and higher temperatures than those measured in the wells. Chemical, isotope, and gas geothermometers indicate a deep fluid temperature of 200-245??C and reflect a relatively short residence time in the fractures feeding the wells. Chloride-enthalpy relations as well as isotopic and chemical compositions of well discharges, thermal springs, and local cold waters support a conceptual model of ascending high-temperature (minimum 225??C) parent fluid that has cooled conductively to form the 160-165??C shallow (to 680 m) fluid encountered by the wells. The hot-spring waters are formed by boiling and steam loss from more or less conductively cooled parent fluid. The more dilute boiling spring waters (Cl = ???32 mg/kg) have cooled from > 225??C to about 160??C by conduction and from 160??C to 98??C by boiling. The most concentrated boiling spring waters (Cl = 37 mg/kg) have cooled from > 225??C to about 200??C by conduction and from 200??C to 98??C by boiling. Intermediate concentrations reflect mixed cooling paths. ?? 1991.

Water absorption characteristic of interlocking compressed earth brick units

NASA Astrophysics Data System (ADS)

Bakar, B. H. Abu; Saari, S.; Surip, N. A.

2017-10-01

This study aims to investigate the water absorption characteristic of interlocking compressed earth brick (ICEB) units. Apart from compressive strength, water absorption is an important property in masonry. This property can affect the quality of the brick itself and the bond strength between the brick and mortar in masonry structures and can result in reducing its strength properties. The units were tested for 24 h water absorption and 5 h boiling water absorption. A total of 170 ICEB units from four ICEB types underwent both tests. For the 24 h water absorption, the ICEB units were dried in the oven for 24 h and then cooled before being weighed. Thereafter, each brick was immersed in water for 24 h and weighed. The same specimens used for the 24 h water absorption test were re-used for the 5 h boiling water absorption test. After completing the 24 h water absorption test, the brick was boiled for 5-hours and weighed. The highest water absorption for the ICEBs in the 24-hour water absorption and 5 h boiling water absorption tests are 15.09% and 17.18%, respectively. The half brick has the highest water absorption (15.87%), whereas the beam brick has the lowest (13.20%). The water absorption of an ICEB unit is higher than that of normal bricks, although the water absorption of the former remains below the maximum rate of the brick water absorption (21%).

Reasons for the lack of chemical stability of treated water rich in magnesium.

PubMed

Swietlik, Joanna; Raczyk-Stanisławiak, Urszula; Piszora, Paweł; Nawrocki, Jacek

2011-12-01

Chemical stability of water should be high enough to ensure that the water reaching the consumers would have the same composition as at the treatment plant. The drinking water supplied by one of the water treatment plants for the city of Poznań was observed to produce periodically white non-sedimenting precipitate on boiling, deteriorating its organoleptic properties. The phenomenon was found to be related to a high content of magnesium in the water taken for treatment and low content of other ions besides bicarbonates. XRD and SEM analyses have shown that a low ratio of calcium ions to magnesium ions leads to formation of calcite crystals on water boiling in which a fraction of cationic crystallographic sites are substituted with Mg(2+) ions giving (Ca(1-x)Mg(x))CO(3) crystallites. Such crystallites have smaller size than those of calcite formed on boiling water coming from other Poznań suppliers. The smaller size of the crystallites is responsible for their slower sedimentation and hence the observed increase in the water turbidity on its boiling. It has been proved that the appearance of precipitates in drinking water at the consumers can be achieved by reduction of the Mg/(Mg + Ca) ratio to below 3, which would inhibit peptisation of the precipitate and prevent water opacity and/or adjustment of pH of the raw water and removal of the carbon dioxide released to convert some carbonate hardness into non-carbonate one. These measures will limit the amount of the precipitate forming upon water boiling and change its microcrystalline type into an easier sedimenting one. Copyright © 2011 Elsevier Ltd. All rights reserved.

Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F.

Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimatedmore » costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP.« less

Effect of soaking, boiling, and steaming on total phenolic contentand antioxidant activities of cool season food legumes.

PubMed

Xu, Baojun; Chang, Sam K C

2008-09-01

The effects of soaking, boiling and steaming processes on the total phenolic components and antioxidant activity in commonly consumed cool season food legumes (CSFL's), including green pea, yellow pea, chickpea and lentil were investigated. As compared to original unprocessed legumes, all processing steps caused significant (p<0.05) decreases in total phenolic content (TPC), DPPH free radical scavenging activity (DPPH) in all tested CSFL's. All soaking and atmospheric boiling treatments caused significant (p<0.05) decreases in oxygen radical absorbing capacity (ORAC). However, pressure boiling and pressure steaming caused significant (p<0.05) increases in ORAC values. Steaming treatments resulted in a greater retention of TPC, DPPH, and ORAC values in all tested CSFL's as compared to boiling treatments. To obtain cooked legumes with similar palatability and firmness, pressure boiling shortened processing time as compared to atmospheric boiling, resulted in insignificant differences in TPC, DPPH for green and yellow pea. However, TPC and DPPH in cooked lentils differed significantly between atmospheric and pressure boiling. As compared to atmospheric processes, pressure processes significantly increased ORAC values in both boiled and steamed CSFL's. Greater TPC, DPPH and ORAC values were detected in boiling water than that in soaking and steaming water. Boiling also caused more solid loss than steaming. Steam processing exhibited several advantages in retaining the integrity of the legume appearance and texture of the cooked product, shortening process time, and greater retention of antioxidant components and activities. Copyright © 2008 Elsevier Ltd. All rights reserved.

Community challenges when using large plastic bottles for Solar Energy Disinfection of Water (SODIS).

PubMed

Borde, Preeti; Elmusharaf, Khalifa; McGuigan, Kevin G; Keogh, Michael B

2016-09-05

Communities living in developing countries as well as populations affected by natural or man-made disasters can be left at great risk from water related diseases, especially those spread through the faecal-oral route. Conventional water treatments such as boiling and chlorination can be effective but may prove costly for impoverished communities. Solar water disinfection (SODIS) has been shown to be a cheap and effective way for communities to treat their water. The exposure to sunlight is typically carried out in small volume plastic beverage bottles (up to 2 l). Given the water requirements of consumption and basic personal hygiene, this may not always meet the needs of communities. Recent work has shown 19-L plastic water dispenser containers to be effective SODIS reactors, comparable in efficacy to PET bottles. In this paper we outline the need for studying SODIS in large volumes and discuss 4 main associated challenges. Apart from clean water needed for consumption, access to adequate water is essential for sanitation and hygiene. Contamination of treated water through unwashed hands or vessels contributes heavily to the spread of water borne pathogens in communities. Traditional water treatments such as boiling and chlorination can be effective but may prove financially burdensome for low income communities. SODIS in large vessels could be used as a simple method to meet water requirements in low income and disaster affected populations. However, there have been some concerns associated with the conventional SODIS method; we identify the main ones to be: (1) cold or cloudy weather; (2) the fear of leaching in plastic bottles; (3) water turbidity, and; (4) community acceptance. The application of SODIS in large bottles like WDCs has the potential to be an efficient and cost effective method of disinfecting water, either for consumption until more rigorous water treatments can be put in place, or for sanitation and hygiene to curb the spread of fecal contamination. Further research is needed that can address some of the limitations and challenges associated with the use of large bottles for SODIS.

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

[Fluorescence spectra analysis of the scrophularia soup].

PubMed

Yan, Li-hua; Song, Feng; Han, Juan; Su, Jing; Qu, Fei-fei; Song, Yi-zhan; Hu, Bo-lin; Tian, Jian-guo

2008-08-01

The cold-water and boiled-water soaked scrophularia soups have been prepared. The emission and excitation spectra of each scrophularia soup under different conditions have been measured at room temperature. The pH values of the different scrophularia soups have been also detected. There are obvious differences between the cold-water soaked scrophularia soup and the boiled-water soaked scrophularia. For both soups the emission wavelength increases with the wavelength of the excitation, but the peaks of the emission spectra for cold-water and boiled-water soaked scrophularia soup are different, which are 441 and 532 nm, respectively. Excitation spectrum has double peaks in the cold-water soaked scrophularia soup while only one peak with longer wavelength in the boiled-water soaked one. The pH value changes from 5.5 to 4.1. According to the organic admixture fluorescence mechanism we analyzed the reasons of the experimental results. Through heating, the interaction in different fluorescence molecular and the energy transfer process in the same fluorescence molecular become more active, and the conjugate structures and the generation of hydrogen bonds, increase. The fluorescence measurement is of value for the scrophularia pharmacology analysis and provides an analytical method for the quality identification of scrophularia soup.

Rapid depressurization event analysis in BWR/6 using RELAP5 and contain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueftueoglu, A.K.; Feltus, M.A.

1995-09-01

Noncondensable gases may become dissolved in Boiling Water Reactor (BWR) water level instrumentation during normal operations. Any dissolved noncondensable gases inside these water columns may come out of solution during rapid depressurization events, and displace water from the reference leg piping resulting in a false high level. These water level errors may cause a delay or failure in actuation, or premature shutdown of the Emergency Core Cooling System. (ECCS). If a rapid depressurization causes an erroneously high water level, preventing automatic ECCS actuation, it becomes important to determine if there would be other adequate indications for operator response and othermore » signals for automatic actuation such as high drywell pressure. It is also important to determine the effect of the level signal on ECCS operation after it is being actuated. The objective of this study is to determine the detailed coupled containment/NSSS response during this rapid depressurization events in BWR/6. The selected scenarios involve: (a) inadvertent opening of all ADS valves, (b) design basis (DB) large break loss of coolant accident (LOCA), and (c) main steam line break (MSLB). The transient behaviors are evaluated in terms of: (a) vessel pressure and collapsed water level response, (b) specific transient boundary conditions, (e.g., scram, MSIV closure timing, feedwater flow, and break blowdown rates), (c) ECCS initiation timing, (d) impact of operator actions, (e) whether indications besides low-low water level were available. The results of the analysis had shown that there would be signals to actuate ECCS other than low reactor level, such as high drywell pressure, low vessel pressure, high suppression pool temperature, and that the plant operators would have significant indications to actuate ECCS.« less

Edwin I. Hatch nuclear plant implementation of improved technical specifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahler, S.R.; Pendry, D.

1994-12-31

Edwin I. Hatch nuclear plant consists of two General Electric boiling water reactor/4 units, with a common control room and a common refueling floor. In March 1993, Hatch began conversion of both units` technical specifications utilizing NUREG 1433. The technical specifications amendment request was submitted February 25, 1994. Issuance is scheduled for October 21, 1994, with implementation on March 15, 1994. The current unit-1 technical specifications are in the {open_quotes}custom{close_quotes} format, and the unit-2 technical specifications are in the old standard format. Hatch previously relocated the fire protection and radiological technical specifications requirements. The Hatch conversion will provide consistency betweenmore » the two units, to the extent practicable.« less

Performance of a personal neutron dosemeter based on direct ion storage at workplace fields in the nuclear industry.

PubMed

Boschung, M; Fiechtner, A; Wernli, C

2007-01-01

In the framework of the EVIDOS project, funded by the EC, measurements were carried out using dosemeters, based on ionisation chambers with direct ion storage (DIS-N), at several workplace fields, namely, at a fuel processing plant, a boiling and a pressurised water reactor, and near transport and storage casks. The measurements and results obtained with the DIS-N in these workplaces, which are representative for the nuclear industry, are described in this study. Different dosemeter configurations of converter and shielding materials were considered. The results are compared with values for personal dose equivalent which were assessed within the EVIDOS project by other partners. The advantages and limitations of the DIS-N dosemeter are discussed.

Level-2 IPE for the Laguna Verde NPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arellano, J.; De Loera, M.A.; Rea, R.

1996-12-31

In response to generic letter GL 88-20, Comision Federal de Electricidad and Instituto de Investigaciones Electricas have jointly developed the individual plant examination (IPE) for the Laguna Verde nuclear power station unit I (LVNPS). This plant is a 675-MW(electric) boiling water reactor (BWR/5) with a reinforced concrete Mark-II containment. The approach used to fulfill the IPE requirements was to make a level-1 probabilistic risk assessment (IPE level 1) plus a containment performance analysis including the behavior and release of the fission products to the environment (IPE level 2). This paper describes the level-2 portion of the LVNPS IPE, paying specialmore » attention to both some improvements to the traditional analytical methods and to the main results.« less

Safety evaluation report on Tennessee Valley Authority: Browns Ferry nuclear performance plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1989-10-01

This safety evaluation report (SER) on the information submitted by the Tennessee Valley Authority (TVA) in its Nuclear Performance Plan, through Revision 2, for the Browns Ferry Nuclear Plant and in supporting documents has been prepared by the US Nuclear Regulatory commission staff. The Browns Ferry Nuclear Plant consists of three boiling-water reactors at a site in Limestone County, Alabama. The plan addresses the plant-specific concerns requiring resolution before the startup of Unit 2. The staff will inspect implementation of those TVA programs that address these concerns. Where systems are common to Units 1 and 2 or to Units 2more » and 3, the staff safety evaluations of those systems are included herein. 85 refs.« less

Correlational approach to turbulent saturated film boiling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chu, T.Y.

A correlation method for saturated film boiling is proposed. The correlation is based on the analogy between film boiling and natural convection. As in the case of natural convection, the turbulent film boiling correlation takes the form of a Nusselt number versus the Raleigh number power law, Nu[sub B] [proportional to] Ra[sub B][sup 1.3]. The proposed correlation shows very good agreement with current data for film boiling of water from vertical surfaces. The general applicability of the correlation is established by comparisons with film boiling data from R-113 and cryogenic fluids. 25 refs., 8 figs.

Markov Model of Accident Progression at Fukushima Daiichi

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuadra A.; Bari R.; Cheng, L-Y

2012-11-11

On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unitmore » 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.« less

The initiation of boiling during pressure transients. [water boiling on metal surfaces

NASA Technical Reports Server (NTRS)

Weisman, J.; Bussell, G.; Jashnani, I. L.; Hsieh, T.

1973-01-01

The initiation of boiling of water on metal surfaces during pressure transients has been investigated. The data were obtained by a new technique in which light beam fluctuations and a pressure signal were simultaneously recorded on a dual beam oscilloscope. The results obtained agreed with those obtained using high speed photography. It was found that, for water temperatures between 90-150 C, the wall superheat required to initiate boiling during a rapid pressure transient was significantly higher than required when the pressure was slowly reduced. This result is explained by assuming that a finite time is necessary for vapor to fill the cavity at which the bubble originates. Experimental measurements of this time are in reasonably good agreement with calculations based on the proposed theory. The theory includes a new procedure for estimating the coefficient of vaporization.

Transetherification method

DOEpatents

Hearn, D.

1985-04-09

Transetherification is carried out in a catalytic distillation reactor, wherein the catalytic structure also serves as a distillation structure, by feeding a first ether to the catalyst bed to at least partially dissociate it into a first olefin and a first alcohol while concurrently therewith feeding either a second olefin (preferably a tertiary olefin) having a higher boiling point than said first olefin or a second alcohol having a higher boiling point than said first alcohol to the catalyst whereby either the second olefin and the first alcohol or the first olefin and the second alcohol react to form a second ether which has a higher boiling point than the first ether, which second ether is concurrently removed as a bottoms in the concurrent reaction-distillation to force that reaction to completion, while the unreacted first olefin or first alcohol is removed in the overhead. 1 fig.

Transetherification method

DOEpatents

Hearn, Dennis

1985-01-01

Transetherification is carried out in a catalytic distillation reactor, wherein the catalytic structure also serves as a distillation structure, by feeding a first ether to the catalyst bed to at least partially dissociate it into a first olefin and a first alcohol while concurrently therewith feeding either a second olefin (preferably a tertiary olefin) having a higher boiling point than said first olefin or a second alcohol having a higher boiling point than said first alcohol to the catalyst whereby either the second olefin and the first alcohol or the first olefin and the second alcohol react to form a second ether which has a higher boiling point than the first ether, which second ether is concurrently removed as a bottoms in the concurrent reaction-distillation to force that reaction to completion, while the unreacted first olefin or first alcohol is removed in the overhead.

Effect of nanostructure on rapid boiling of water on a hot copper plate: a molecular dynamics study

NASA Astrophysics Data System (ADS)

Fu, Ting; Mao, Yijin; Tang, Yong; Zhang, Yuwen; Yuan, Wei

2016-08-01

Molecular dynamic simulations are performed to study the effects of nanostructure on rapid boiling of water that is suddenly heated by a hot copper plate. The results show that the nanostructure has significant effects on energy transfer from solid copper plate to liquid water and phase change process from liquid water to vapor. The liquid water on the solid surface rapidly boil after contacting with an extremely hot copper plate and consequently a cluster of liquid water moves upward during phase change. The temperature of the water film when it separates from solid surface and its final temperature when the system is at equilibrium strongly depend on the size of the nanostructure. These temperatures increase with increasing size of nanostructure. Furthermore, a non-vaporized molecular layer is formed on the surface of the copper plate even continuous heat flux is passing into water domain through the plate.

Comparison of vapor formation of water at the solid/water interface to colloidal solutions using optically excited gold nanostructures.

PubMed

Baral, Susil; Green, Andrew J; Livshits, Maksim Y; Govorov, Alexander O; Richardson, Hugh H

2014-02-25

The phase transformation properties of liquid water to vapor is characterized by optical excitation of the lithographically fabricated single gold nanowrenches and contrasted to the phase transformation properties of gold nanoparticles located and optically excited in a bulk solution system [two and three dimensions]. The 532 nm continuous wave excitation of a single gold nanowrench results in superheating of the water to the spinodal decomposition temperature of 580 ± 20 K with bubble formation below the spinodal decomposition temperature being a rare event. Between the spinodal decomposition temperature and the boiling point liquid water is trapped into a metastable state because a barrier to vapor nucleation exists that must be overcome before the thermodynamically stable state is realized. The phase transformation for an optically heated single gold nanowrench is different from the phase transformation of optically excited colloidal gold nanoparticles solution where collective heating effects dominates and leads to the boiling of the solution exactly at the boiling point. In the solution case, the optically excited ensemble of nanoparticles collectively raises the ambient temperature of water to the boiling point where liquid is converted into vapor. The striking difference in the boiling properties of the single gold nanowrench and the nanoparticle solution system can be explained in terms of the vapor-nucleation mechanism, the volume of the overheated liquid, and the collective heating effect. The interpretation of the observed regimes of heating and vaporization is consistent with our theoretical modeling. In particular, we explain with our theory why the boiling with the collective heating in a solution requires 3 orders of magnitude less intensity compared to the case of optically driven single nanowrench.

The challenge of improving boiling: lessons learned from a randomized controlled trial of water pasteurization and safe storage in Peru.

PubMed

Heitzinger, K; Rocha, C A; Quick, R E; Montano, S M; Tilley, D H; Mock, C N; Carrasco, A J; Cabrera, R M; Hawes, S E

2016-07-01

Boiling is the most common method of household water treatment in developing countries; however, it is not always effectively practised. We conducted a randomized controlled trial among 210 households to assess the effectiveness of water pasteurization and safe-storage interventions in reducing Escherichia coli contamination of household drinking water in a water-boiling population in rural Peru. Households were randomized to receive either a safe-storage container or a safe-storage container plus water pasteurization indicator or to a control group. During a 13-week follow-up period, households that received a safe-storage container and water pasteurization indicator did not have a significantly different prevalence of stored drinking-water contamination relative to the control group [prevalence ratio (PR) 1·18, 95% confidence interval (CI) 0·92-1·52]. Similarly, receipt of a safe-storage container alone had no effect on prevalence of contamination (PR 1·02, 95% CI 0·79-1·31). Although use of water pasteurization indicators and locally available storage containers did not increase the safety of household drinking water in this study, future research could illuminate factors that facilitate the effective use of these interventions to improve water quality and reduce the risk of waterborne disease in populations that boil drinking water.

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Favalli, Andrea; Vo, D.; Grogan, Brandon R.

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

DOE PAGES

Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

2016-02-26

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

Nanofluids for power engineering: Emergency cooling of overheated heat transfer surfaces

NASA Astrophysics Data System (ADS)

Bondarenko, B. I.; Moraru, V. N.; Sidorenko, S. V.; Komysh, D. V.

2016-07-01

The possibility of emergency cooling of an overheated heat transfer surface using nanofluids in the case of a boiling crisis is explored by means of synchronous recording of changes of main heat transfer parameters of boiling water over time. Two nanofluids are tested, which are derived from a mixture of natural aluminosilicates (AlSi-7) and titanium dioxide (NF-8). It is found that the introduction of a small portions of nanofluid into a boiling coolant (distilled water) in a state of film boiling ( t heater > 500°C) can dramatically decrease the heat transfer surface temperature to 130-150°C, which corresponds to a transition to a safe nucleate boiling regime without affecting the specific heat flux. The fact that this regime is kept for a long time at a specific heat load exceeding the critical heat flux for water and t heater = 125-130°C is particularly important. This makes it possible to prevent a potential accident emergency (heater burnout and failure of the heat exchanger) and to ensure the smooth operation of the equipment.

Comparison of lipids in organs of the starfish Asterias amurensis associated with different treatments

NASA Astrophysics Data System (ADS)

Wang, Qi; Ikegame, Keita; Takahashi, Koretaro; Xue, Changhu; Zhang, Weinong; Wang, Hongxun; Hou, Wenfu; Wang, Yuming

2013-09-01

Lipids were extracted from organs of the starfish Asterias amurensis associated with different treatments (raw-control, boiling and heating), and then analyzed for lipid content, lipid oxidation index, lipid classes and fatty acid composition. Results showed that boiling softened the hard starfish shells, thus facilitating the collection of starfish organs. As compared with raw organs, the boiled organs had lower water content and higher lipid content, possibly due to the loss of water-holding capacity caused by protein denaturation. Both boiling and heating increased the peroxide value (PV), thiobarbituric acid (TBA) value and carbon value (CV) of lipids. Despite slight increases in the content of complex lipids, associated lipid composition had no substantial variations upon boiling and heating. For simple lipids, the content of 1, 2-diglyceride decreased in boiled and heated organs, with free fatty acids observed on thin layer chromatography (TLC). However, neither boiling nor heating significantly changed the fatty acid compositions of simple or complex lipids in starfish organs, suggesting that these two treatments had no significant effects on complex lipids in starfish organs. Together, our results indicated that boiling of starfish soon after capture facilitated the handling and extraction of useful complex lipids consisting of abundant glucosylceramide and eicosapentaenoic acid (EPA)-bounded phospholipids.

Science at Home: Measuring a Thermophysical Property of Water with a Microwave Oven

ERIC Educational Resources Information Center

Levine, Zachary H.

2018-01-01

An attempt to calibrate a conventional oven led to making a measurement of a thermophysical property of water using items found in the author's home. Specifically, the ratio of the energy required to heat water from the melting point to boiling to the energy required to completely boil away the water is found to be 5.7. This may be compared to the…

Crimson Viper, Phase IV: Field Experimentation Venue Report and CV14 Combined Report

DTIC Science & Technology

2014-09-01

Notify Officials and Populous of the following: • Notify locations of water source contamination , warning and recommended actions (i.e. boil water) per...of supplies available (e.g. blankets, tarps, food , etc) including medicines Observations and Feedback Objective 1: Assess effectiveness of the...Populous of the following: • Notify locations of water source contamination , warning and recommended actions (i.e. boil water) per local government 69

Assessment of in situ butanol recovery by vacuum during acetone butanol ethanol (ABE) fermentation

USDA-ARS?s Scientific Manuscript database

Butanol fermentation is product limiting due to butanol toxicity to microbial cells. Butanol (boiling point: 118 deg C) boils at a greater temperature than water (boiling point: 100 deg C) and application of vacuum technology to integrated acetone-butanol-ethanol (ABE) fermentation and recovery may ...

Some methods for achieving more efficient performance of fuel assemblies

NASA Astrophysics Data System (ADS)

Boltenko, E. A.

2014-07-01

More efficient operation of reactor plant fuel assemblies can be achieved through the use of new technical solutions aimed at obtaining more uniform distribution of coolant over the fuel assembly section, more intense heat removal on convex heat-transfer surfaces, and higher values of departure from nucleate boiling ratio (DNBR). Technical solutions using which it is possible to obtain more intense heat removal on convex heat-transfer surfaces and higher DNBR values in reactor plant fuel assemblies are considered. An alternative heat removal arrangement is described using which it is possible to obtain a significantly higher power density in a reactor plant and essentially lower maximal fuel rod temperature.

Experimental study on pool boiling of distilled water and HFE7500 fluid under microgravity

NASA Astrophysics Data System (ADS)

Yang, Yan-jie; Chen, Xiao-qian; Huang, Yi-yong; Li, Guang-yu

2018-02-01

The experimental study on bubble behavior and heat transfer of pool boiling for distilled water and HFE7500 fluid under microgravity has been conducted by using drop tower in the National Microgravity Laboratory of China (NMLC). Two MCH ceramic plates of 20 mm(L) × 10 mm(W) × 1.2 mm(H) were used as the heaters. The nucleate boiling evolution under microgravity was observed during the experiment. It has been found that at the same heat flux, the bubbles of HFE7500 (which has smaller contact angle) grew faster and bigger, moved quickly on the heater surface, and were easier to merge into a central big bubble with other bubbles than that of distilled water. The whole process of bubbles coalescence from seven to one was recorded by using video camera. For distilled water (with bigger contact angle), the bubbles tended to keep at the nucleate location on heater surface, and the central big bubble evolved at its nucleate cite by absorbing smaller bubbles nearby. Compared with the bubbles under normal gravity, bubble radius of distilled water under microgravity was about 1.4 times bigger and of HFE7500 was about more than 6 times bigger till the end of experiment. At the beginning, pool boiling heat transfer of distilled water was advanced and then impeded under microgravity. As to HFE7500, the pool boiling impedes the heat transfer from heater to liquid under microgravity throughout the experiment.

Daily variations in effluent water turbidity and diarrhoeal illness in a Russian city.

PubMed

Egorov, Andrey I; Naumova, Elena N; Tereschenko, Andrey A; Kislitsin, Victor A; Ford, Timothy E

2003-03-01

To assess an association between temporal variations in drinking water quality and gastrointestinal (GI) illness, a cohort study involving 100 randomly selected families (367 individuals) was conducted in the city of Cherepovets, Russia from June through November 1999. Participants maintained daily diaries of gastrointestinal symptoms, water consumption and other behavioural exposure variables, while daily effluent water quality data were provided by the water utility. The cumulative incidence rate of self-reported gastrointestinal diseases, 1.7 cases per person-year, was almost two orders of magnitude higher than that of officially reported GI infections in the city. An interquartile range increase in effluent water turbidity of 0.8 Nephelometric Turbidity Units was associated with a relative risk of self-reported GI illness of 1.47 (95% Confidence Interval 1.16, 1.86) at a lag of 2 days after control for daily rate of consumption of non-boiled tap water, behavioural covariates, day of the week and a seasonally-related linear trend. In the analysis by subsets of study participants stratified by non-boiled tap water consumption, no statistically significant associations between turbidity and GI illness were found for the study participants who always boiled their drinking water. For individuals who drank non-boiled tap water, statistically significant associations between turbidity and GI illness were detected at lags 1, 2 and 7 days.

Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Douglas W. Akers; Edwin A. Harvego

2012-08-01

This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coble, Jamie; Orton, Christopher; Schwantes, Jon

Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test threemore » fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.« less

Measurement of Key Pool BOiling Parameters in nanofluids for Nuclerar Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bang, In C; Buongiorno, Jdacopo; Hu, Lin-wen

Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e.,more » contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.« less

Electrochemical study of aluminum corrosion in boiling high purity water

NASA Technical Reports Server (NTRS)

Draley, J. E.; Legault, R. A.

1969-01-01

Electrochemical study of aluminum corrosion in boiling high-purity water includes an equation relating current and electrochemical potential derived on the basis of a physical model of the corrosion process. The work involved an examination of the cathodic polarization behavior of 1100 aluminum during aqueous oxidation.

Activity ratios in soil contaminated by the source of different reactor condition in the FDNPP accident

NASA Astrophysics Data System (ADS)

Satou, Yukihiko; Sueki, Keisuke; Sasa, Kimikazu; Matsunaka, Tetsuya; Shibayama, Nao; Takahashi, Tsutomu; Kinoshita, Norikazu

2014-05-01

The Fukushima Dai-ichi Nuclear power plant (FDNPP) accident caused radioactive contamination on the surface soil at Fukushima and its adjacent prefectures. Substantial contamination has been found in the northwestern area from the FDNPP, according to the airborne monitoring and ground base survey by the Japanese government. Activity ratios would have characteristic information on emission sources because each relevant reactor had different amount of radionuclide and different activity ratio. The ratios can be used to clarify more detailed source and process in the contamination. We have addressed to consider them in Namie town, northwestern region from the FDNPP. This study focused on the gamma-ray emitting radionuclides of 134Cs, 137Cs, and 110mAg. The activities were decay-corrected as of 11th March, 2011 when all nuclear reactors scrammed. Data of activity ratios by our results and the Japanese official report classified the investigated northwestern region into 3 groups. Ratios of 0.02 for 110mAg/137Cs and 0.90 for 134Cs/137Cs were observed in the northern region of 15 km inside from the FDNPP. On the other hand, two kinds of 110mAg/137Cs ratios of 0.005 and 0.002 were distributed broadly in the region 60 km away from the plant. The 134Cs/137Cs ratio was 0.98 there. The activity ratios of 110mAg/137Cs and 134Cs/137Cs in the northern region from the FDNPP correspond to those of nuclear fuel in Unit 1 according to estimation using the ORIGEN code. The 134Cs/137Cs in the northwestern area from FDNPP agrees with that of Unit 2 and 3. The 110mAg/137Cs ratios of 0.005 and0.002 are 1/5 - 1/10 of the Unit 2 and 3. Official report has announced that discharges of the radionuclides from Unit 2 and 3 occurred on 14th March, 2011. It is known that contamination in the northwestern region from the FDNPP took place on 15th March, 2011. Plausible species for silver in reactor core, metal, and halide etc. have higher boiling point than those species for cesium. The core would be cooled down to lower temperature of the boiling point of silver at the timing contamination occurred. Thus, silver with higher boiling point was not much released than cesium with lower boiling point. The 110mAg/137Cs ratio has served to identify the specific sources of contamination in the northwestern area from the FDNPP.

Bottom head to shell junction assembly for a boiling water nuclear reactor

DOEpatents

Fife, Alex Blair; Ballas, Gary J.

1998-01-01

A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening.

Condensation model for the ESBWR passive condensers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

Bottom head to shell junction assembly for a boiling water nuclear reactor

DOEpatents

Fife, A.B.; Ballas, G.J.

1998-02-24

A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input thatmore » describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.« less

Water NSTF Design, Instrumentation, and Test Planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Darius D.; Gerardi, Craig D.; Hu, Rui

The following report serves as a formal introduction to the water-based Natural convection Shutdown heat removal Test Facility (NSTF) program at Argonne. Since 2005, this US Department of Energy (DOE) sponsored program has conducted large scale experimental testing to generate high-quality and traceable validation data for guiding design decisions of the Reactor Cavity Cooling System (RCCS) concept for advanced reactor designs. The most recent facility iteration, and focus of this report, is the operation of a 1/2 scale model of a water-RCCS concept. Several features of the NSTF prototype align with the conceptual design that has been publicly released formore » the AREVA 625 MWt SC-HTGR. The design of the NSTF also retains all aspects common to a fundamental boiling water thermosiphon, and thus is well poised to provide necessary experimental data to advance basic understanding of natural circulation phenomena and contribute to computer code validation. Overall, the NSTF program operates to support the DOE vision of aiding US vendors in design choices of future reactor concepts, advancing the maturity of codes for licensing, and ultimately developing safe and reliable reactor technologies. In this report, the top-level program objectives, testing requirements, and unique considerations for the water cooled test assembly are discussed, and presented in sufficient depth to support defining the program’s overall scope and purpose. A discussion of the proposed 6-year testing program is then introduced, which outlines the specific strategy and testing plan for facility operations. The proposed testing plan has been developed to meet the toplevel objective of conducting high-quality test operations that span across a broad range of single- and two-phase operating conditions. Details of characterization, baseline test cases, accident scenario, and parametric variations are provided, including discussions of later-stage test cases that examine the influence of geometric variations and off-normal configurations. The facility design follows, including as-built dimensions and specifications of the various mechanical and liquid systems, design choices for the test section, water storage tank, and network piping. Specifications of the instrumentation suite are then presented, along with specific information on performance windows, measurement uncertainties, and installation locations. Finally, descriptions of the control systems and heat removal networks are provided, which have been engineered to support precise quantification of energy balances and facilitate well-controlled test operations.« less

US Efforts in Support of Examinations at Fukushima Daiichi – 2016 Evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amway, P.; Andrews, N.; Bixby, Willis

Although it is clear that the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limitedmore » full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings (TEPCO) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document reports recent results from the US Forensics Effort to use information obtained by TEPCO to enhance the safety of existing and future nuclear power plant designs. This Forensics Effort, which is sponsored by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of US experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO information from Daiichi that address these needs. Examples presented in this report demonstrate that significant safety insights are being obtained in the areas of component performance, fission product release and transport, debris end-state location, and combustible gas generation and transport. In addition to reducing uncertainties related to severe accident modeling progression, these insights are being used to update guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, reduced uncertainties in modeling the events at Daiichi will improve the realism of reactor safety evaluations and inform future D&D activities by improving the capability for characterizing potential hazards to workers involved with cleanup activities.« less

Generation and evolution of hydrothermal fluids at Yellowstone: Insights from the Heart Lake Geyser Basin

NASA Astrophysics Data System (ADS)

Lowenstern, J. B.; Bergfeld, D.; Evans, W. C.; Hurwitz, S.

2012-01-01

We sampled fumaroles and hot springs from the Heart Lake Geyser Basin (HLGB), measured water and gas discharge, and estimated heat and mass flux from this geothermal area in 2009. The combined data set reveals that diverse fluids share an origin by mixing of deep solute-rich parent water with dilute heated meteoric water, accompanied by subsequent boiling. A variety of chemical and isotopic geothermometers are consistent with a parent water that equilibrates with rocks at 205°C ± 10°C and then undergoes 21% ± 2% adiabatic boiling. Measured diffuse CO2 flux and fumarole compositions are consistent with an initial dissolved CO2 concentration of 21 ± 7 mmol upon arrival at the caldera boundary and prior to southeast flow, boiling, and discharge along the Witch Creek drainage. The calculated advective flow from the basin is 78 ± 16 L s-1of parent thermal water, corresponding to 68 ± 14 MW, or ˜1% of the estimated thermal flux from Yellowstone. Helium and carbon isotopes reveal minor addition of locally derived crustal, biogenic, and meteoric gases as this fluid boils and degasses, reducing the He isotope ratio (Rc/Ra) from 2.91 to 1.09. The HLGB is one of the few thermal areas at Yellowstone that approaches a closed system, where a series of progressively boiled waters can be sampled along with related steam and noncondensable gas. At other Yellowstone locations, steam and gas are found without associated neutral Cl waters (e.g., Hot Spring Basin) or Cl-rich waters emerge without significant associated steam and gas (Upper Geyser Basin).

Generation and evolution of hydrothermal fluids at Yellowstone: Insights from the Heart Lake Geyser Basin

USGS Publications Warehouse

Lowenstern, J. B.; Bergfeld, D.; Evans, William C.; Hurwitz, S.

2012-01-01

We sampled fumaroles and hot springs from the Heart Lake Geyser Basin (HLGB), measured water and gas discharge, and estimated heat and mass flux from this geothermal area in 2009. The combined data set reveals that diverse fluids share an origin by mixing of deep solute-rich parent water with dilute heated meteoric water, accompanied by subsequent boiling. A variety of chemical and isotopic geothermometers are consistent with a parent water that equilibrates with rocks at 205°C ± 10°C and then undergoes 21% ± 2% adiabatic boiling. Measured diffuse CO2 flux and fumarole compositions are consistent with an initial dissolved CO2 concentration of 21 ± 7 mmol upon arrival at the caldera boundary and prior to southeast flow, boiling, and discharge along the Witch Creek drainage. The calculated advective flow from the basin is 78 ± 16 L s−1 of parent thermal water, corresponding to 68 ± 14 MW, or –1% of the estimated thermal flux from Yellowstone. Helium and carbon isotopes reveal minor addition of locally derived crustal, biogenic, and meteoric gases as this fluid boils and degasses, reducing the He isotope ratio (Rc/Ra) from 2.91 to 1.09. The HLGB is one of the few thermal areas at Yellowstone that approaches a closed system, where a series of progressively boiled waters can be sampled along with related steam and noncondensable gas. At other Yellowstone locations, steam and gas are found without associated neutral Cl waters (e.g., Hot Spring Basin) or Cl-rich waters emerge without significant associated steam and gas (Upper Geyser Basin).

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent researchmore » progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.« less

Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

NASA Astrophysics Data System (ADS)

Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

2016-09-01

Past decades have seen active research in enhancement of boiling heat transfer by surface modifications. Favorable surface modifications are expected to enhance boiling efficiency. Several interrelated mechanisms such as capillarity, surface energy alteration, wettability, cavity geometry, wetting transitions, geometrical features of surface morphology, etc., are responsible for change in the boiling behavior of modified surfaces. Not much work is available on pool boiling at low pressures on microscale/nanoscale geometries; low pressure boiling is attractive in many applications wherein low operating temperatures are desired for a particular working fluid. In this background, an experimental setup was designed and developed to investigate the pool boiling performance of water on (a) plain aluminum micro wire (99.999 % pure) and, (b) nano-porous alumina structured aluminum micro wire, both having diameter of 250 µm, under sub-atmospheric pressure. Nano-structuring on the plain wire surface was achieved via anodization. Two samples, A and B of anodized wires, differing by the degree of anodization were tested. The heater length scale (wire diameter) was much smaller than the capillary length scale. Pool boiling characteristics of water were investigated at three different sub-atmospheric pressures of 73, 123 and 199 mbar (corresponding to T sat = 40, 50 and 60 °C). First, the boiling characteristics of plain wire were measured. It was noticed that at sub-atmospheric pressures, boiling heat transfer performance for plain wire was quite low due to the increased bubble sizes and low nucleation site density. Subsequently, boiling performance of nano-structured wires (both Sample A and Sample B) was compared with plain wire and it was noted that boiling heat transfer for the former was considerably enhanced as compared to the plain wire. This enhancement is attributed to increased nucleation site density, change in wettability and possibly due to enhanced pore scale evaporation. A preliminary estimation of the bubble growth rates, measured by high speed videography, was undertaken and compared with classical bubble growth rate correlations. It was observed that the average bubble departure sizes on Sample B were larger as compared to plain wire, due to larger surface forces holding the bubble before departure. Bubble condensation in the thermal boundary layer was also captured.

Catalytic two-stage coal hydrogenation and hydroconversion process

DOEpatents

MacArthur, James B.; McLean, Joseph B.; Comolli, Alfred G.

1989-01-01

A process for two-stage catalytic hydrogenation and liquefaction of coal to produce increased yields of low-boiling hydrocarbon liquid and gas products. In the process, the particulate coal is slurried with a process-derived liquid solvent and fed at temperature below about 650.degree. F. into a first stage catalytic reaction zone operated at conditions which promote controlled rate liquefaction of the coal, while simultaneously hydrogenating the hydrocarbon recycle oils at conditions favoring hydrogenation reactions. The first stage reactor is maintained at 650.degree.-800.degree. F. temperature, 1000-4000 psig hydrogen partial pressure, and 10-60 lb coal/hr/ft.sup.3 reactor space velocity. The partially hydrogenated material from the first stage reaction zone is passed directly to the close-coupled second stage catalytic reaction zone maintained at a temperature at least about 25.degree. F. higher than for the first stage reactor and within a range of 750.degree.-875.degree. F. temperature for further hydrogenation and thermal hydroconversion reactions. By this process, the coal feed is successively catalytically hydrogenated and hydroconverted at selected conditions, which results in significantly increased yields of desirable low-boiling hydrocarbon liquid products and minimal production of undesirable residuum and unconverted coal and hydrocarbon gases, with use of less energy to obtain the low molecular weight products, while catalyst life is substantially increased.

75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-22

... March 2, 2010, at 11545 Rockville Pike, Rockville, Maryland, Room T2 B3. The meeting will be open to... chapters (1, 4, 11, 12, 15 and 18) of the Safety Evaluation Report with Open Items associated with the STP...

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The New York Power Authority (NYPA) and Entergy Corp. of New Orleans, La., announced recently the signing of a memorandum of understanding as a step toward a contract for Entergy to provide management services to NYPA`s two nuclear power plants. The agreement is the first of its kind. NYPA is the nation`s largest state-owned electric utility and supplier of one-quarter of New York`s electricity. Its nuclear plants are Indian Point 3 (IP3) in Buchanan, Westchester County, and James A. FitzPatrick in Scriba, Oswego County. Entergy is a utility holding company and its subsidiary, Entergy Operations Inc., is widely recognized asmore » one of the leading nuclear operators in the United States. {open_quotes}NYPA`s nuclear plants are assets that belong to the people of New York,{close_quotes} said C.D. {open_quotes}Rapp{close_quotes} Rappleyea, NYPA`s chairman and CEO. {open_quotes}Our alliance with Entergy can provide the people of this state with added assurance that these plants will operate with the highest level of safety and efficiency.{close_quotes} FitzPatrick, an 800 MW boiling water reactor, has operated since 1975 and IP3, a 980 MW pressurized water reactor, since 1976. Although both are currently running well, they have had problems in recent years, and IP3 is on the US Nuclear Regulatory Commission`s (NRC) list of plants requiring increased regulatory attention. Entergy operated both types of reactors, has three single-unit sites like NYPA`s and is experienced in operating plants for different utility owners.« less

Round Robin Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Klamerus, E.W.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. Several organizations from the US, Europe, and Asia were invited to participate in a Round Robin analysis to perform independent pretest predictions and posttest evaluations of the behavior of the SCV model during the high pressure test. Both pretest and posttest analysis results from all Round Robin participants were compared tomore » the high pressure test data. This paper summarizes the Round Robin analysis activities and discusses the lessons learned from the collective effort.« less

Characterization of cracking behavior using posttest fractographic analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobayashi, T.; Shockey, D.A.

A determination of time to initiation of stress corrosion cracking in structures and test specimens is important for performing structural failure analysis and for setting inspection intervals. Yet it is seldom possible to establish how much of a component's lifetime represents the time to initiation of fracture and how much represents postinitiation crack growth. This exploratory research project was undertaken to examine the feasibility of determining crack initiation times and crack growth rates from posttest examination of fracture surfaces of constant-extension-rate-test (CERT) specimens by using the fracture reconstruction applying surface topography analysis (FRASTA) technique. The specimens used in this studymore » were Type 304 stainless steel fractured in several boiling water reactor (BWR) aqueous environments. 2 refs., 25 figs., 2 tabs.« less

The first PANDA tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dreier, J.; Huggenberger, M.; Aubert, C.

1996-08-01

The PANDA test facility at PSI in Switzerland is used to study the long-term Simplified Boiling Water Reactor (SBWR) Passive Containment Cooling System (PCCS) performance. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensables in the system. The PANDA facility has a 1:1 vertical scale, and 1:25 ``system`` scale (volume, power, etc.). Steady-state PCCS condenser performance tests and extensive facility characterization tests have been completed. Transient system behavior tests were conducted late in 1995; results from the first three transient tests (M3 series) aremore » reviewed. The first PANDA tests showed that the overall global behavior of the SBWR containment was globally repeatable and very favorable; the system exhibited great ``robustness.``« less

Fuel Performance Calculations for FeCrAl Cladding in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents themore » analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.« less

Key Parameters for Operator Diagnosis of BWR Plant Condition during a Severe Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clayton, Dwight A.; Poore, III, Willis P.

2015-01-01

The objective of this research is to examine the key information needed from nuclear power plant instrumentation to guide severe accident management and mitigation for boiling water reactor (BWR) designs (specifically, a BWR/4-Mark I), estimate environmental conditions that the instrumentation will experience during a severe accident, and identify potential gaps in existing instrumentation that may require further research and development. This report notes the key parameters that instrumentation needs to measure to help operators respond to severe accidents. A follow-up report will assess severe accident environmental conditions as estimated by severe accident simulation model analysis for a specific US BWR/4-Markmore » I plant for those instrumentation systems considered most important for accident management purposes.« less

Watching the Pot to Improve Inquiry Skills

ERIC Educational Resources Information Center

Brockway, Debra; Papaleo, Marie

2009-01-01

The International Boiling Point Project is an online, collaborative project for students in grades 6-12 in which they investigate the impact of four factors (room temperature, elevation, volume of water and heating device) on the boiling point of water. A detailed procedure is provided for students in order to limit the number of variables…

Effect of CrO3 Sealing Time on Anodized A12024-T3

NASA Astrophysics Data System (ADS)

Korda, Akhmad A.; Hidayat, R. Z.

2016-08-01

The effect of CrO3 sealing time on anodized aluminum alloy has been investigated. A1 2024-T3 were used as substrate. Anodizing was carried out using chromic acid. CrO3 sealing was conducted in CrO3 solution for 30, 60, 90, 120 and 150 minutes. As comparison, other specimens were also prepared as anodized and boiled water sealing. Thickness of the coating was observed by optical microscope. Anodized and sealing layer was analyzed by X- ray diffraction. The hardness of as anodized, boiled water sealing and CrO3 sealing were compared. The highest hardness is achieved by CrO3 sealed specimen and followed by boiled water sealing and as anodized specimens. The longer the processes of CrO3 sealing the higher layer thickness and therefore the higher hardness of the oxide layer. The best resistance to electrolyte penetration is achieved by the CrO3 sealed specimen followed by boiled water sealed and as anodized specimens. The higher thickness of oxide layer, the higher the resistance against electrolyte penetration.

Effects of Hull Scratching, Soaking, and Boiling on Antinutrients in Japanese Red Sword Bean (Canavalia gladiata).

PubMed

Une, Satsuki; Nonaka, Koji; Akiyama, Junich

2016-10-01

The effects of hull processing, soaking, and boiling on the content or activity of antinutrients in the red sword bean (RSB; Canavalia gladiata) were investigated. RSB seeds were compared with kidney bean (KB; Phaseolus vulgaris) seeds that are starch based and often used as processed products in Japan. RSB seeds had higher weight, thicker hull, and higher protein content, but lower moisture content compared with KB seeds. Because of the strong and thick hull, the relative water absorption of untreated RSB seeds was very low after soaking. Seeds were soaked after dehulling, scratching, and roasting. The results showed that hull scratching was the optimal method for increasing water absorption during soaking compared with dehulling and roasting. After soaking, the water used for soaking was discarded, since it had a high content of polyphenols and bitter taste, and RSB seeds were boiled in fresh water for 20, 40, and 60 min. The results showed that polyphenol and tannin contents, antioxidant activity, and hemagglutinating activity, as well as maltase, sucrase, and trypsin inhibitor activities in scratched RSB seeds decreased significantly after boiling compared with those in raw seeds, whereas amylase inhibitor activity showed no significant change. Overall, it was concluded that the combination of hull scratching, soaking, and boiling in fresh water can reduce thermal-stable or sensitive antinutrients in RSB and thus, significantly improve its nutritional value. © 2016 Institute of Food Technologists®.

A scaling law for the local CHF on the external bottom side of a fully submerged reactor vessel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

1997-02-01

A scaling law for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water has been developed from the results of an advanced hydrodynamic CHF model for pool boiling on a downward facing curved heating surface. The scaling law accounts for the effects of the size of the vessel, the level of liquid subcooling, the intrinsic properties of the fluid, and the spatial variation of the local critical heat flux along the heating surface. It is found that for vessels with diameters considerably larger than the characteristic size ofmore » the vapor masses, the size effect on the local critical heat flux is limited almost entirely to the effect of subcooling associated with the local liquid head. When the subcooling effect is accounted for separately, the local CHF limit is nearly independent of the vessel size. Based upon the scaling law developed in this work, it is possible to merge, within the experimental uncertainties, all the available local CHF data obtained for various vessel sizes under both saturated and subcooled boiling conditions into a single curve. Applications of the scaling law to commercial-size vessels have been made for various system pressures and water levels above the heated vessel. Over the range of conditions explored in this study, the local CHF limit is found to increase by a factor of two or more from the bottom center to the upper edge of the vessel. Meanwhile, the critical heat flux at a given angular position of the heated vessel is also found to increase appreciably with the system pressure and the water level.« less

Laboratory Production of Biofuels and Biochemicals from a Rapeseed Oil through Catalytic Cracking Conversion.

PubMed

Ng, Siauw H; Shi, Yu; Heshka, Nicole E; Zhang, Yi; Little, Edward

2016-09-02

The work is based on a reported study which investigates the processability of canola oil (bio-feed) in the presence of bitumen-derived heavy gas oil (HGO) for production of transportation fuels through a fluid catalytic cracking (FCC) route. Cracking experiments are performed with a fully automated reaction unit at a fixed weight hourly space velocity (WHSV) of 8 hr(-1), 490-530 °C, and catalyst/oil ratios of 4-12 g/g. When a feed is in contact with catalyst in the fluid-bed reactor, cracking takes place generating gaseous, liquid, and solid products. The vapor produced is condensed and collected in a liquid receiver at -15 °C. The non-condensable effluent is first directed to a vessel and is sent, after homogenization, to an on-line gas chromatograph (GC) for refinery gas analysis. The coke deposited on the catalyst is determined in situ by burning the spent catalyst in air at high temperatures. Levels of CO2 are measured quantitatively via an infrared (IR) cell, and are converted to coke yield. Liquid samples in the receivers are analyzed by GC for simulated distillation to determine the amounts in different boiling ranges, i.e., IBP-221 °C (gasoline), 221-343 °C (light cycle oil), and 343 °C+ (heavy cycle oil). Cracking of a feed containing canola oil generates water, which appears at the bottom of a liquid receiver and on its inner wall. Recovery of water on the wall is achieved through washing with methanol followed by Karl Fischer titration for water content. Basic results reported include conversion (the portion of the feed converted to gas and liquid product with a boiling point below 221 °C, coke, and water, if present) and yields of dry gas (H2-C2's, CO, and CO2), liquefied petroleum gas (C3-C4), gasoline, light cycle oil, heavy cycle oil, coke, and water, if present.

Cooking under Pressure: Applying the Ideal Gas Law in the Kitchen

ERIC Educational Resources Information Center

Chen, Ling; Anderson, Jennifer Y.; Wang, Diane R.

2010-01-01

This case study uses a daily cooking scenario to demonstrate how the boiling point of water is directly related to the external pressures in order to reinforce the concepts of boiling and boiling point, apply ideal gas law, and relate chemical reaction rates with temperatures. It also extends its teaching to autoclaves used to destroy…

Nutrition content of brisket point end of part Simental Ongole Crossbred meat in boiled various temperature

NASA Astrophysics Data System (ADS)

Riyanto, J.; Sudibya; Cahyadi, M.; Aji, A. P.

2018-01-01

This aim of this study was to determine the quality of nutritional contents of beef brisket point end of Simental Ongole Crossbred meat in various boiling temperatures. Simental Ongole Crossbred had been fattened for 9 months. Furthermore, they were slaughtered at slaughterhouse and brisket point end part of meat had been prepared to analyse its nutritional contents using Food Scan. These samples were then boiled at 100°C for 0 (TR), 15 (R15), and 30 (R30) minutes, respectively. The data was analysed using Randomized Complete Design (CRD) and Duncan’s multiple range test (DMRT) had been conducted to differentiate among three treatments. The results showed that boiling temperatures significantly affected moisture, and cholesterol contents of beef (P<0.05) while fat content was not significantly affected by boiling temperatures. The boiling temperature decreased beef water contents from 72.77 to 70.84%, on the other hand, the treatment increased beef protein and cholesterol contents from 20.77 to 25.14% and 47.55 to 50.45 mg/100g samples, respectively. The conclusion of this study was boiling of beef at 100°C for 15 minutes and 30 minutes decreasing water content and increasing protein and cholesterol contents of brisket point end of Simental Ongole Crossbred beef.

Modeling of Kerena Emergency Condenser

NASA Astrophysics Data System (ADS)

Bryk, Rafał; Schmidt, Holger; Mull, Thomas; Wagner, Thomas; Ganzmann, Ingo; Herbst, Oliver

2017-12-01

KERENA is an innovative boiling water reactor concept equipped with several passive safety systems. For the experimental verification of performance of the systems and for codes validation, the Integral Test Stand Karlstein (INKA) was built in Karlstein, Germany. The emergency condenser (EC) system transfers heat from the reactor pressure vessel (RPV) to the core flooding pool in case of water level decrease in the RPV. EC is composed of a large number of slightly inclined tubes. During accident conditions, steam enters into the tubes and condenses due to the contact of the tubes with cold water at the secondary side. The condensed water flows then back to the RPV due to gravity. In this paper two approaches for modeling of condensation in slightly inclined tubes are compared and verified against experiments. The first approach is based on the flow regime map. Depending on the regime, heat transfer coefficient is calculated according to specific semi-empirical correlation. The second approach uses a general, fully-empirical correlation. The models are developed with utilization of the object-oriented Modelica language and the open-source OpenModelica environment. The results are compared with data obtained during a large scale integral test, simulating loss of coolant accident performed at Integral Test Stand Karlstein (INKA). The comparison shows a good agreement.Due to the modularity of models, both of them may be used in the future in systems incorporating condensation in horizontal or slightly inclined tubes. Depending on his preferences, the modeller may choose one-equation based approach or more sophisticated model composed of several exchangeable semi-empirical correlations.

Tunable integration of absorption-membrane-adsorption for efficiently separating low boiling gas mixtures near normal temperature

PubMed Central

Liu, Huang; Pan, Yong; Liu, Bei; Sun, Changyu; Guo, Ping; Gao, Xueteng; Yang, Lanying; Ma, Qinglan; Chen, Guangjin

2016-01-01

Separation of low boiling gas mixtures is widely concerned in process industries. Now their separations heavily rely upon energy-intensive cryogenic processes. Here, we report a pseudo-absorption process for separating low boiling gas mixtures near normal temperature. In this process, absorption-membrane-adsorption is integrated by suspending suitable porous ZIF material in suitable solvent and forming selectively permeable liquid membrane around ZIF particles. Green solvents like water and glycol were used to form ZIF-8 slurry and tune the permeability of liquid membrane surrounding ZIF-8 particles. We found glycol molecules form tighter membrane while water molecules form looser membrane because of the hydrophobicity of ZIF-8. When using mixing solvents composed of glycol and water, the permeability of liquid membrane becomes tunable. It is shown that ZIF-8/water slurry always manifests remarkable higher separation selectivity than solid ZIF-8 and it could be tuned to further enhance the capture of light hydrocarbons by adding suitable quantity of glycol to water. Because of its lower viscosity and higher sorption/desorption rate, tunable ZIF-8/water-glycol slurry could be readily used as liquid absorbent to separate different kinds of low boiling gas mixtures by applying a multistage separation process in one traditional absorption tower, especially for the capture of light hydrocarbons. PMID:26892255

Pool Boiling Experiment Has Five Successful Flights

NASA Technical Reports Server (NTRS)

Chiaramonte, Fran

1997-01-01

The Pool Boiling Experiment (PBE) is designed to improve understanding of the fundamental mechanisms that constitute nucleate pool boiling. Nucleate pool boiling is a process wherein a stagnant pool of liquid is in contact with a surface that can supply heat to the liquid. If the liquid absorbs enough heat, a vapor bubble can be formed. This process occurs when a pot of water boils. On Earth, gravity tends to remove the vapor bubble from the heating surface because it is dominated by buoyant convection. In the orbiting space shuttle, however, buoyant convection has much less of an effect because the forces of gravity are very small. The Pool Boiling Experiment was initiated to provide insight into this nucleate boiling process, which has many earthbound applications in steamgeneration power plants, petroleum plants, and other chemical plants. In addition, by using the test fluid R-113, the Pool Boiling Experiment can provide some basic understanding of the boiling behavior of cryogenic fluids without the large cost of an experiment using an actual cryogen.

Water evaporation in silica colloidal deposits.

PubMed

Peixinho, Jorge; Lefèvre, Grégory; Coudert, François-Xavier; Hurisse, Olivier

2013-10-15

The results of an experimental study on the evaporation and boiling of water confined in the pores of deposits made of mono-dispersed silica colloidal micro-spheres are reported. The deposits are studied using scanning electron microscopy, adsorption of nitrogen, and adsorption of water through attenuated total reflection-infrared spectroscopy. The evaporation is characterized using differential scanning calorimetry and thermal gravimetric analysis. Optical microscopy is used to observe the patterns on the deposits after evaporation. When heating at a constant rate and above boiling temperature, the release of water out of the deposits is a two step process. The first step is due to the evaporation and boiling of the surrounding and bulk water and the second is due to the desorption of water from the pores. Additional experiments on the evaporation of water from membranes having cylindrical pores and of heptane from silica deposits suggest that the second step is due to the morphology of the deposits. Copyright © 2013 Elsevier Inc. All rights reserved.

An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwang-Won, Lee; Sang-Yong, Lee

1995-09-01

A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by themore » frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.« less

The Effect of Different Boiling and Filtering Devices on the Concentration of Disinfection By-Products in Tap Water

PubMed Central

Carrasco-Turigas, Glòria; Villanueva, Cristina M.; Goñi, Fernando; Rantakokko, Panu; Nieuwenhuijsen, Mark J.

2013-01-01

Disinfection by-products (DBPs) are ubiquitous contaminants in tap drinking water with the potential to produce adverse health effects. Filtering and boiling tap water can lead to changes in the DBP concentrations and modify the exposure through ingestion. Changes in the concentration of 4 individual trihalomethanes (THM4) (chloroform (TCM), bromodichloromethane (BDCM), dibromochloromethane (DBCM), and bromoform (TBM)), MX, and bromate were tested when boiling and filtering high bromine-containing tap water from Barcelona. For filtering, we used a pitcher-type filter and a household reverse osmosis filter; for boiling, an electric kettle, a saucepan, and a microwave were used. Samples were taken before and after each treatment to determine the change in the DBP concentration. pH, conductivity, and free/total chlorine were also measured. A large decrease of THM4 (from 48% to 97%) and MX concentrations was observed for all experiments. Bromine-containing trihalomethanes were mostly eliminated when filtering while chloroform when boiling. There was a large decrease in the concentration of bromate with reverse osmosis, but there was a little effect in the other experiments. These findings suggest that the exposure to THM4 and MX through ingestion is reduced when using these household appliances, while the decrease of bromate is device dependent. This needs to be considered in the exposure assessment of the epidemiological studies. PMID:23476675

The effect of ginger and garlic addition during cooking on the volatile profile of grass carp (Ctenopharyngodon idella) soup.

PubMed

Li, Jin-Lin; Tu, Zong-Cai; Zhang, Lu; Sha, Xiao-Mei; Wang, Hui; Pang, Juan-Juan; Tang, Ping-Ping

2016-08-01

Ginger and garlic have long been used in Asian countries to enhance the flavor and to neutralize any unpleasant odors present in fish soup. The purpose of this study was to evaluate the change in the amount of volatile components present in fish soup compared to boiled water solutions of ginger and garlic. The fish soup was prepared by boiling oil-fried grass carp ( Ctenopharyngodon idella ) with or without ginger and/or garlic. Generally, boiling garlic and ginger in water led to a decrease in the amount of the principal volatile constituents of these spices, together with the formation of some new volatiles such as pentanal, hexanal, and nonanal. The results showed that 16 terpenes present in raw ginger, predominantly camphene, β -phellandrene, β -citral, α -zingiberene, and ( E )-neral, were detected in fish soup with added ginger and thus remained in the solution even after boiling. Similarly, 2-propen-1-ol and three sulfur compounds (allyl sulfide, diallyl disulfide, and diallyl trisulfide) present in raw garlic, were present in trace amounts in the boiled garlic solution, but were present in considerably larger amounts in the boiled fish solution with garlic or garlic plus ginger. In conclusion, the effect of adding spices on the volatile profile of grass carp soup can be attributed to the dissolution of flavor volatiles mainly derived from raw spices into the solution, with few additional volatiles being formed during boiling. In addition, boiling previously fried grass carp with spices led to enhanced volatile levels compared to boiled spice solutions.

Transesterification of rapeseed oil for biodiesel production in trickle-bed reactors packed with heterogeneous Ca/Al composite oxide-based alkaline catalyst.

PubMed

Meng, Yong-Lu; Tian, Song-Jiang; Li, Shu-Fen; Wang, Bo-Yang; Zhang, Min-Hua

2013-05-01

A conventional trickle bed reactor and its modified type both packed with Ca/Al composite oxide-based alkaline catalysts were studied for biodiesel production by transesterification of rapeseed oil and methanol. The effects of the methanol usage and oil flow rate on the FAME yield were investigated under the normal pressure and methanol boiling state. The oil flow rate had a significant effect on the FAME yield for the both reactors. The modified trickle bed reactor kept over 94.5% FAME yield under 0.6 mL/min oil flow rate and 91 mL catalyst bed volume, showing a much higher conversion and operational stability than the conventional type. With the modified trickle bed reactor, both transesterification and methanol separation could be performed simultaneously, and glycerin and methyl esters were separated additionally by gravity separation. Copyright © 2013 Elsevier Ltd. All rights reserved.

Boiling water scarification plus stratification improves germination of Iliamna rivularis (Malvaceae) seeds

Treesearch

Katri Himanen; Markku Nygren; R. Kasten Dumroese

2012-01-01

Scarification with boiling water plus stratification was most effective in improving germination of Iliamna rivularis (Douglas ex Hook.) Greene (Malvaceae) in an experiment that compared 3 treatments. Seeds from 15 sites representing 5 western US states were used in the experiment. Initial response of the seedlots to the treatments was similar, apart from one seedlot....

Water preparation practices in south Kalimantan, Indonesia.

PubMed

Prihartono, N; Adisasmita, A; Costello, C; Damayanti, R; Prasetyo, S; Syarif, S

1994-12-01

This study investigated water preparation practices, water sources, and sanitation measures in 400 households in four villages in Banjar district of South Kalimantan, Indonesia. A survey of randomly selected households in each of the four villages was undertaken during the season of low prevalence of diarrhoeal disease (February-March 1989). A follow-up survey of the same households was conducted 6 months later to measure prevalence during the peak season. Twenty-three in-depth interviews were carried out with key persons in local villages. Observations on behaviour in the local food stalls were recorded. The results of the study indicate that 97% of the households report that they regularly boil their drinking water. However, 37% of the households regularly or occasionally mix boiled with unboiled water for drinking, or use unboiled water alone. The mixing of boiled with unboiled water is particularly frequent in the preparation of 'cold tea', a popular drink in households and food stalls. The occasional or regular use of unboiled water varies by ethnicity, education and literacy, and economic status. The use of unboiled water is associated with higher rates of childhood diarrhoeas in the households studied.

Pourbaix diagrams for the ternary system of iron-chromium-nickel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beverskog, B.; Puigdomenech, I.

1999-11-01

Pourbaix diagrams (potential-pH diagrams) for the ternary system of Fe-Cr-Ni at 25 C to 300 C were calculated. Extrapolation of thermochemical data to elevated temperatures was performed with the revised model of Helgeson-Kirkham-Flowers, which also allows uncharged aqueous complexes to be handled. The large stability of the bimetallic spinel oxides (trevorite [NiFe{sub 2}O{sub 4}], chromite [FeCr{sub 2}O{sub 4}], and nichromite [NiCr{sub 2}O{sub 4}]) is shown by their predominance areas on top of those for the single metal Pourbaix diagrams. NiFe{sub 2}O{sub 4} had the largest stability area of the spinels, and it covered the entire potential range for the stabilitymore » of water at intermediate pH. FeCr{sub 2}O{sub 4} had the smallest stability area and was the least stable of the bimetallic spinels. Results were discussed in connection with the different chemistries used in nuclear power reactors of the boiling water type.« less

Extremophiles and the search for extraterrestrial life.

PubMed

Cavicchioli, Ricardo

2002-01-01

Extremophiles thrive in ice, boiling water, acid, the water core of nuclear reactors, salt crystals, and toxic waste and in a range of other extreme habitats that were previously thought to be inhospitable for life. Extremophiles include representatives of all three domains (Bacteria, Archaea, and Eucarya); however, the majority are microorganisms, and a high proportion of these are Archaea. Knowledge of extremophile habitats is expanding the number and types of extraterrestrial locations that may be targeted for exploration. In addition, contemporary biological studies are being fueled by the increasing availability of genome sequences and associated functional studies of extremophiles. This is leading to the identification of new biomarkers, an accurate assessment of cellular evolution, insight into the ability of microorganisms to survive in meteorites and during periods of global extinction, and knowledge of how to process and examine environmental samples to detect viable life forms. This paper evaluates extremophiles and extreme environments in the context of astrobiology and the search for extraterrestrial life.

Effects of Swirler Shape on Two-Phase Swirling Flow in a Steam Separator

NASA Astrophysics Data System (ADS)

Kataoka, Hironobu; Shinkai, Yusuke; Tomiyama, Akio

Experiments on two-phase swirling flow in a separator are carried out using several swirlers having different vane angles, different hub diameters and different number of vanes to seek a way for improving steam separators of uprated boiling water reactors. Ratios of the separated liquid flow rate to the total liquid flow rate, flow patterns, liquid film thicknesses and pressure drops are measured to examine the effects of swirler shape on air-water two-phase swirling annular flows in a one-fifth scale model of the separator. As a result, the following conclusions are obtained for the tested swirlers: (1) swirler shape scarcely affects the pressure drop in the barrel of the separator, (2) decreasing the vane angle is an effective way for reducing the pressure drop in the diffuser of the separator, and (3) the film thickness at the inlet of the pick-off-ring of the separator is not sensitive to swirler shape, which explains the reason why the separator performance does not depend on swirler shape.

Development of a fully-consistent reduced order model to study instabilities in boiling water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dykin, V.; Demaziere, C.

2012-07-01

A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to othermore » existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C{sub mn}{sup *V,D} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate Eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities. (authors)« less

Distribution of Trichloroethylene and Geologic Controls on Contaminant Pathways near the Royal River, McKin Superfund Site Area, Gray, Maine

USGS Publications Warehouse

Lyford, Forest P.; Flight, L.E.; Stone, Janet Radway; Clifford, Scott

1999-01-01

Vapor-diffusion samplers were used in the autumn of 1997 to determine the lateral extent and distribution of concentrations of a trichloroethylene (TCE) plume in the ground-water discharge area near the McKin Superfund Site, Gray, Maine. Analyses of vapor in the samplers identified a plume about 800 feet wide entering the river near Boiling Springs, an area of ground-water discharge on the flood plain of the Royal River. The highest observed concentration of TCE in vapor was in an area of sand boils on the western bank of the river and about 200 feet downstream from Boiling Springs. Previous studies showed that most of the TCE load in the river originated in the area of the sand boils. In general, highest concentrations were observed on the western side of the river on the upgradient side of the plume, but TCE also was detected at numerous locations in the center and eastern bank of the river. The TCE plume discharges to the river where fine-grained glaciomarine sediments of the Presumpscot Formation are absent and where coarse-grained facies of buried glaciomarine fan deposits provide a pathway for ground-water flow. Based on results of analyses of vapor-diffusion samples and other previous studies, the plume appears to pass under and beyond the river near Boiling Springs and along the river for about 300 feet downstream from the sand boils. A coarse-grained, organic-rich layer at the base of the alluvial flood plain sediments is confined by overlying fine-grained alluvial sediments and may provide a conduit for ground-water leaking upward from buried glaciomarine fan deposits.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Comparison of carbon monoxide levels during heating of ice and water to boiling point with a camping stove.

PubMed

Leigh-Smith, Simon; Watt, Ian; McFadyen, Angus; Grant, Stan

2004-01-01

To determine whether using a camping stove to bring a pan of ice to boiling point produces higher carbon monoxide (CO) concentration than would bringing a pan of water to boiling point. The hypothesis was that ice would cause greater CO concentration because of its greater flame-cooling effect and, consequently, more incomplete combustion. This was a randomized, prospective observational study. After an initial pilot study, CO concentration was monitored during 10 trials for each of ice and water. A partially ventilated 200-L cardboard box model was developed and then used inside a chamber at -6 degrees C. Ice temperature and volume, water temperature and volume, pan size, and flame characteristics were all standardized. Temperature of the heated medium was monitored to determine time to boiling point. Carbon monoxide concentration was monitored every 30 seconds for the first 3 minutes, then every minute until the end of each 10-minute trial. There was no significant difference (P > .05) in CO production levels between ice and water. Each achieved a similar mean plateau level of approximately 400 ppm CO concentration with a similar rate of rise. However, significantly higher (P = .014) CO concentration occurred at 4 and 5 minutes when the flame underwent a yellow flare; this occurred only on 3 occasions when ice was the medium. There were no significant differences for CO production between bringing a pan of ice or water to boiling point. In a small number of ice trials, the presence of a yellow flame resulted in high CO concentration. Yellow flares might occur more often with ice or snow melting, but this has not been proven.

El Cobreloa: A geyser with two distinct eruption styles

NASA Astrophysics Data System (ADS)

Namiki, A.; Munoz, C.; Manga, M.; Hurwitz, S.; King, E.; Negri, A.; Ortega, P.; Patel, A.; Rudolph, M.

2013-12-01

El Cobreloa geyser has two distinct eruption styles: vigorous major eruptions, and less energetic minor eruptions. Minor eruptions splash hot water intermittently over an approximately 4 minute time period. Major eruptions begin with an eruption style similar to minor eruptions, but then transition to a voluminous and water-dominated eruption, and finally end with energetic steam discharge. The steam discharge continues for approximately 1 hour. We calculated the eruption intervals by visual observations, acoustic measurements, and ground temperature measurements. All of measurements consistently show that each eruption style has a regular interval: 4 hours and 40 minutes for major eruptions, and ~13 minutes for minor eruptions. From these observations, we infer that there are two boiling loci that source each type of eruption, one at the bottom and the other at the top of the conduit. If the bottom of the conduit is hot enough, boiling begins at the bottom of the conduit to make a steam slug. As this slug ascends in the conduit, it heats the surrounding water. If the slug rises fast enough it splashes water when it reaches the surface, creating minor eruptions. Each successive steam slug continues to heat water in the conduit until it eventually reaches the boiling temperature everywhere. Once the top of the conduit begins boiling, the energetic steam discharge begins and the boiling propagates downward. Such a process causes major eruption. Geysers are often studied as an analogue to magmatic volcanoes because it is easier to document how mass and energy transfer lead to eruptions. El Cobreloa provides insight into how the system becomes primed for large eruptions.

Transport synthetic acceleration for long-characteristics assembly-level transport problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zika, M.R.; Adams, M.L.

2000-02-01

The authors apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, the authors take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. The main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme. The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. The authorsmore » devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, they define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. They implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; they prove that the long-characteristics discretization yields an SPD matrix. They present results of the acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly.« less

The Parable of the Boiled System Safety Professional: Drift to Failure

NASA Technical Reports Server (NTRS)

Shivers, C. Herbert

2011-01-01

Recall from the Parable of the Boiled Frog, that tossing a frog into boiling water causes the frog to jump out and hop away while placing a frog in suitable temperature water and slowly bringing the water to a boil results in the frog boiling due to not being aware of the slowly increasing danger, theoretically, of course. System safety professionals must guard against allowing dangers to creep unnoticed into their projects and be ever alert to notice signs of impending problems. People have used various phrases related to the idea, most notably, latent conditions, James Reason in Managing the Risks of Organizational Accidents (1, pp 10-11), Drift to Failure, Sydney Dekker (2, pp 82-86) in Resilience Engineering: Chronicling the Emergence of Confused Consensus in Resilience Engineering: Concepts and Precepts, Hollnagel, Woods and Leveson, and normalization of deviance, Diane Vaughan in The Challenger Launch Decision: Risky Technology, Culture, and Deviance at NASA (3). Reason also said, If eternal vigilance is the price of liberty, then chronic unease is the price of safety (1, p 37). Our challenge as system safety professionals is to be aware of the emergence of signals that warn us of slowly eroding safety margins. This paper will discuss how system safety professionals might better perform in that regard.

Melting Metal on a Playing Card

ERIC Educational Resources Information Center

Greenslade, Thomas B., Jr.

2016-01-01

Many of us are familiar with the demonstration of boiling water in a paper cup held over a candle or a Bunsen burner; the ignition temperature of paper is above the temperature of 100°C at which water boils under standard conditions. A more dramatic demonstration is melting tin held in a playing card. This illustration is from Tissandier's book on…

Bioactive compounds and antioxidant activity of fresh and processed white cauliflower.

PubMed

Ahmed, Fouad A; Ali, Rehab F M

2013-01-01

Brassica species are very rich in health-promoting phytochemicals, including phenolic compounds, vitamin C, and minerals. The objective of this study was to investigate the effect of different blanching (i.e., water and steam) and cooking (i.e., water boiling, steam boiling, microwaving, and stir-frying) methods on the nutrient components, phytochemical contents (i.e., polyphenols, carotenoids, flavonoid, and ascorbic acid), antioxidant activity measured by DPPH assay, and phenolic profiles of white cauliflower. Results showed that water boiling and water blanching processes had a great effect on the nutrient components and caused significant losses of dry matter, protein, and mineral and phytochemical contents. However, steam treatments (blanching and cooking), stir-frying, and microwaving presented the lowest reductions. Methanolic extract of fresh cauliflower had significantly the highest antioxidant activity (68.91%) followed by the extracts of steam-blanched, steam-boiled, stir-fried, and microwaved cauliflower 61.83%, 59.15%, 58.93%, and 58.24%, respectively. HPLC analysis revealed that the predominant phenolics of raw cauliflower were protocatechuic acid (192.45), quercetin (202.4), pyrogallol (18.9), vanillic acid (11.90), coumaric acid (6.94), and kaempferol (25.91) mg/100 g DW, respectively.

Bioactive Compounds and Antioxidant Activity of Fresh and Processed White Cauliflower

PubMed Central

Ahmed, Fouad A.; Ali, Rehab F. M.

2013-01-01

Brassica species are very rich in health-promoting phytochemicals, including phenolic compounds, vitamin C, and minerals. The objective of this study was to investigate the effect of different blanching (i.e., water and steam) and cooking (i.e., water boiling, steam boiling, microwaving, and stir-frying) methods on the nutrient components, phytochemical contents (i.e., polyphenols, carotenoids, flavonoid, and ascorbic acid), antioxidant activity measured by DPPH assay, and phenolic profiles of white cauliflower. Results showed that water boiling and water blanching processes had a great effect on the nutrient components and caused significant losses of dry matter, protein, and mineral and phytochemical contents. However, steam treatments (blanching and cooking), stir-frying, and microwaving presented the lowest reductions. Methanolic extract of fresh cauliflower had significantly the highest antioxidant activity (68.91%) followed by the extracts of steam-blanched, steam-boiled, stir-fried, and microwaved cauliflower 61.83%, 59.15%, 58.93%, and 58.24%, respectively. HPLC analysis revealed that the predominant phenolics of raw cauliflower were protocatechuic acid (192.45), quercetin (202.4), pyrogallol (18.9), vanillic acid (11.90), coumaric acid (6.94), and kaempferol (25.91) mg/100 g DW, respectively. PMID:24171164

Unorthodox bubbles when boiling in cold water.

PubMed

Parker, Scott; Granick, Steve

2014-01-01

High-speed movies are taken when bubbles grow at gold surfaces heated spotwise with a near-infrared laser beam heating water below the boiling point (60-70 °C) with heating powers spanning the range from very low to so high that water fails to rewet the surface after bubbles detach. Roughly half the bubbles are conventional: They grow symmetrically through evaporation until buoyancy lifts them away. Others have unorthodox shapes and appear to contribute disproportionately to heat transfer efficiency: mushroom cloud shapes, violently explosive bubbles, and cavitation events, probably stimulated by a combination of superheating, convection, turbulence, and surface dewetting during the initial bubble growth. Moreover, bubbles often follow one another in complex sequences, often beginning with an unorthodox bubble that stirs the water, followed by several conventional bubbles. This large dataset is analyzed and discussed with emphasis on how explosive phenomena such as cavitation induce discrepancies from classical expectations about boiling.

Influence of the heater material on the critical heat load at boiling of liquids on surfaces with different sizes

NASA Astrophysics Data System (ADS)

Anokhina, E. V.

2010-05-01

Data on critical heat loads q cr for the saturated and unsaturated pool boiling of water and ethanol under atmospheric pressure are reported. It is found experimentally that the critical heat load does not necessarily coincide with the heat load causing burnout of the heater, which should be taken into account. The absolute values of q cr for the boiling of water and ethanol on copper surfaces 65, 80, 100, 120, and 200 μm in diameter; tungsten surface 100 μm in diameter; and nichrome surface 100 μm in diameter are obtained experimentally.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

KERENA safety concept in the context of the Fukushima accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

AN EVALUATION OF HEAVY WATER REACTORS FOR POWER

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1957-10-01

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

Does the public receive and adhere to boil water advisory recommendations? A cross-sectional study in Newfoundland and Labrador, Canada.

PubMed

Jones-Bitton, Andria; Gustafson, Diana L; Butt, Kelly; Majowicz, Shannon E

2016-01-05

Highly publicized water supply problems highlight the importance of safe drinking water to the public. Boil water advisories (BWAs) are an important precautionary measure meant to protect public health by ensuring drinking water safety. Newfoundland and Labrador, Canada is a prime location for exploring public notification practices and adherence to recommendations as there were a total of 215 BWAs, affecting 6 % of the provincial population, in 145 communities between April 2006 and March 2007 when data for the present study were collected. Residents who received household water from a public water supply were randomly selected for a telephone interview. Collected data included participants' notification of boil water advisory, satisfaction with information provided, and their adherence to recommendations. Most participants learned that a BWA had been issued or lifted in their community through radio, television, or word of mouth. BWAs were issued for a range of operational reasons. Almost all participants who had experienced a BWA reported wanting more information about the reasons a BWA had been issued. Low adherence to water use recommendations during a BWA was common. This study is first to report on public adherence to boil water advisory recommendations in Canada. The findings raise public health concerns, particularly given the high number of BWAs issued each year. Further studies in partnership with community stakeholders and government decision-makers responsible for overseeing public water systems are needed to assess the perceptions of BWAs, the reasons for non-adherence, and to identify information dissemination methods to increase information uptake and public adherence with acceptable uses of public drinking water during a BWA.

High heat flux burnout in subcooled flow boiling

NASA Astrophysics Data System (ADS)

Celata, G. P.; Cumo, M.; Mariani, A.

1995-09-01

The paper reports the results of an experimental research carried out at the Heat Transfer Division of the Energy Department, C.R. Casaccia, on the thermal hydraulic characterization of subcooled flow boiling CHF under typical conditions of thermonuclear fusion reactors, i.e. high liquid velocity and subcooling. The experiment was carried out exploring the following parameters: channel diameter (from 2.5 to 8.0 mm), heated length (10 and 15 cm), liquid velocity (from 2 to 40 m/s), exit pressure (from atmospheric to 5.0 MPa), inlet temperature (from 30 to 80 °C), channel orientation (vertical and horizontal). A maximum CHF value of 60.6 MW/m2 has been obtained under the following conditions: T in=30°, p=2.5 MPa, u=40 m/s, D=2.5 mm (smooth channel) Turbulence promoters (helically coiled wires) have been employed to further enhance the CHF attainable with subcooled flow boiling. Helically coiled wires allow an increase of 50% of the maximum CHF obtained with smooth channels.

Occurrence of thyroid hormone activities in drinking water from eastern China: contributions of phthalate esters.

PubMed

Shi, Wei; Hu, Xinxin; Zhang, Fengxian; Hu, Guanjiu; Hao, Yingqun; Zhang, Xiaowei; Liu, Hongling; Wei, Si; Wang, Xinru; Giesy, John P; Yu, Hongxia

2012-02-07

Thyroid hormone is essential for the development of humans. However, some synthetic chemicals with thyroid disrupting potentials are detectable in drinking water. This study investigated the presence of thyroid active chemicals and their toxicity potential in drinking water from five cities in eastern China by use of an in vitro CV-1 cell-based reporter gene assay. Waters were examined from several phases of drinking water processing, including source water, finished water from waterworks, tap water, and boiled tap water. To identify the responsible compounds, concentrations and toxic equivalents of a list of phthalate esters were quantitatively determined. None of the extracts exhibited thyroid receptor (TR) agonist activity. Most of the water samples exhibited TR antagonistic activities. None of the boiled water displayed the TR antagonistic activity. Dibutyl phthalate accounted for 84.0-98.1% of the antagonist equivalents in water sources, while diisobutyl phthalate, di-n-octyl phthalate and di-2-ethylhexyl phthalate also contributed. Approximately 90% of phthalate esters and TR antagonistic activities were removable by waterworks treatment processes, including filtration, coagulation, aerobic biodegradation, chlorination, and ozonation. Boiling water effectively removed phthalate esters from tap water. Thus, this process was recommended to local residents to reduce certain potential thyroid related risks through drinking water.

The PANDA tests for SBWR certification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varadi, G.; Dreier, J.; Bandurski, Th.

1996-03-01

The ALPHA project is centered around the experimental and analytical investigation of the long-term decay heat removal from the containments of the next generation of {open_quotes}passive{close_quotes} ALWRs. The project includes integral system tests in the large-scale (1:25 in volume) PANDA facility as well as several other series of tests and supporting analytical work. The first series of experiments to be conducted in PANDA have become a required experimental element in the certification process for the General Electric Simplified Boiling Water Reactor (SBWR). The PANDA general experimental philosophy, facility design, scaling, and instrumentation are described. Steady-state PCCS condenser performance tests andmore » extensive facility characterization tests were already conducted. The transient system behavior tests are underway; preliminary results from the first transient test M3 are reviewed.« less

Benchmarking study of the MCNP code against cold critical experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, S.

1991-01-01

The purpose of this study was to benchmark the widely used Monte Carlo code MCNP against a set of cold critical experiments with a view to using the code as a means of independently verifying the performance of faster but less accurate Monte Carlo and deterministic codes. The experiments simulated consisted of both fast and thermal criticals as well as fuel in a variety of chemical forms. A standard set of benchmark cold critical experiments was modeled. These included the two fast experiments, GODIVA and JEZEBEL, the TRX metallic uranium thermal experiments, the Babcock and Wilcox oxide and mixed oxidemore » experiments, and the Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory (PNL) nitrate solution experiments. The principal case studied was a small critical experiment that was performed with boiling water reactor bundles.« less

Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, William BJ J

2016-01-01

A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blademore » histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.« less

Vacuum breaker valve assembly

DOEpatents

Thompson, J.L.; Upton, H.A.

1999-04-27

Breaker valve assemblies for a simplified boiling water nuclear reactor are described. The breaker valve assembly, in one form, includes a valve body and a breaker valve. The valve body includes an interior chamber, and an inlet passage extends from the chamber and through an inlet opening to facilitate transporting particles from outside of the valve body to the interior chamber. The breaker valve is positioned in the chamber and is configured to substantially seal the inlet opening. Particularly, the breaker valve includes a disk which is sized to cover the inlet opening. The disk is movably coupled to the valve body and is configured to move substantially concentrically with respect to the valve opening between a first position, where the disk completely covers the inlet opening, and a second position, where the disk does not completely cover the inlet opening. 1 fig.

Vacuum breaker valve assembly

DOEpatents

Thompson, Jeffrey L.; Upton, Hubert Allen

1999-04-27

Breaker valve assemblies for a simplified boiling water nuclear reactor are described. The breaker valve assembly, in one form, includes a valve body and a breaker valve. The valve body includes an interior chamber, and an inlet passage extends from the chamber and through an inlet opening to facilitate transporting particles from outside of the valve body to the interior chamber. The breaker valve is positioned in the chamber and is configured to substantially seal the inlet opening. Particularly, the breaker valve includes a disk which is sized to cover the inlet opening. The disk is movably coupled to the valve body and is configured to move substantially concentrically with respect to the valve opening between a first position, where the disk completely covers the inlet opening, and a second position, where the disk does not completely cover the inlet opening.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walter, Matthew; Yin, Shengjun; Stevens, Gary

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-temperaturemore » (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.« less

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Effect of boiling, frying, and baking on recovery of aflatoxin from naturally contaminated corn grits or cornmeal.

PubMed

Stoloff, L; Trucksess, M W

1981-05-01

Corn grits naturally contaminated with aflatoxins were used for making boiled grits, and portions of the boiled grits were used for making pan-fried grits; cornmeal naturally contaminated with aflatoxins was used for making corn muffins. Procedures and recipes were derived from cookbook and market package recommendations. From analyses of the products for aflatoxins before and after preparation of the table-ready products, it was determined that 72 +/- 9% (n = 15) of the aflatoxin found in the original grits could be recovered after the grits were boiled. The recovery of aflatoxin B1 after the grits were fried was either 66 +/- 10% (n = 6) or 47 +/- 8% (n = 9), depending on whether 3 cups of water or 4 cups of water per cup of grits, respectively, were used for preparing the boiled grits before frying. Similarly, it was determined that 87 +/- 4% (n = 9) of the aflatoxin B1 found in the original cornmeal could be recovered from the baked muffins. No detectable aflatoxin B2 a was present in the extracts from any of the table-ready products.

Recent Work on Flow Boiling and Condensation in a Single Microchannel

NASA Astrophysics Data System (ADS)

Quan, Xiaojun; Wang, Guodong; Cheng, Ping; Wu, Huiying

2007-06-01

Recent visualization and measurements results on flow boiling of water and condensation of steam in a single microchannel, carried out at Shanghai Jiaotong University, is summarized in this paper. For flow boiling of water, experiments were conducted in a single microchannel with a trapezoidal cross-section having a hydraulic diameter of 186 μm and a length of 30 mm. A boiling flow pattern map in terms of heat flux versus mass flux, showing the unstable and stable boiling flow regimes in the microchannel, is obtained. For the investigation of condensation, experiments were carried out for steam condensing inside a single microchannel with a length of 60mm having a hydraulic diameter of 87 μm and 120μm respectively. The location of transition from annular flow to plug/slug flow in a microchannel is found to be dependent on both the dimensionless condensation heat transfer rate as well as the Reynolds number of the steam. The frequency for the occurrence of the injection flow is found to increase with the increasing mass flux.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Evaporation of a sessile water drop and a drop of aqueous salt solution.

PubMed

Misyura, S Y

2017-11-07

The influence of various factors on the evaporation of drops of water and aqueous salt solution has been experimentally studied. Typically, in the studies of drop evaporation, only the diffusive vapor transfer, radiation and the molecular heat conduction are taken into account. However, vapor-gas convection plays an important role at droplet evaporation. In the absence of droplet boiling, the influence of gas convection turns out to be the prevailing factor. At nucleate boiling, a prevailing role is played by bubbles generation and vapor jet discharge at a bubble collapse. The gas convection behavior for water and aqueous salt solution is substantially different. With a growth of salt concentration over time, the influence of the convective component first increases, reaches an extremum and then significantly decreases. At nucleate boiling in a salt solution it is incorrect to simulate the droplet evaporation and the heat transfer in quasi-stationary approximation. The evaporation at nucleate boiling in a liquid drop is divided into several characteristic time intervals. Each of these intervals is characterized by a noticeable change in both the evaporation rate and the convection role.

Production of superheated steam from vapor-dominated geothermal reservoirs

USGS Publications Warehouse

Truesdell, A.H.; White, D.E.

1973-01-01

Vapor-dominated geothermal systems such as Larderello, Italy, The Geysers, California, and Matsukawa, Japan yield dry or superheated steam when exploited. Models for these systems are examined along with production data and the thermodynamic properties of water, steam and rock. It is concluded that these systems initially consist of a water and steam filled reservoir, a water-saturated cap rock, and a water or brine-saturated deep reservoir below a water table. Most liquid water in all parts of the system is relatively immobilized in small pores and crevices; steam dominates the large fractures and voids of the reservoir and is the continuous, pressure-controlling phase. With production, the pressure is lowered and the liquid water boils, causing massive transfer of heat from the rock and its eventual drying. Passage of steam through already dried rock produces superheating. After an initial vaporization of liquid water in the reservoir, the decrease in pressure produces increased boiling below the deep water table. With heavy exploitation, boiling extends deeper into hotter rock and the temperature of the steam increases. This model explains most features of the published production behavior of these systems and can be used to guide exploitation policies. ?? 1973.

The effect of post-mortem ageing and heating on water retention in bovine muscles.

PubMed

Kołczak, Tadeusz; Krzysztoforski, Krzysztof; Palka, Krystyna

2007-04-01

The muscles semitendinosus (ST) and psoas major (PM) were removed from chilled young bull carcasses 24h after slaughter and stored at 4°C. At the 1st, 6th and 12th day of post-mortem ageing the chemical composition (moisture, fat, protein, collagen) and contents of free, immobilized and unfreezable water in the muscles were estimated. The muscle steaks were boiled at 100°C, roasted at 170°C or fried at 160°C to an internal temperature of 75°C, and the amounts of total, free, immobilized, and unfreezable water in heated muscles were evaluated. The unfreezable water was estimated by DSC. In the raw muscles immobilized water constituted 74-75%, free water 16.6-17.6% and unfreezable water 7-8% of the total water. Independent of time of ageing, PM muscle contained significantly more free water than ST muscle. During post-mortem ageing, changes in free, immobilized and unfreezable water in muscles were not significant. The level of free water was highest in boiled and least in fried meat, however the amount of immobilized water was highest in fried and lowest in boiled meat. The amount of unfreezable water in muscles heated after 12 days of post-mortem ageing decreased.

A citation-based assessment of the performance of U.S. boiling water reactors following extended power up-rates

NASA Astrophysics Data System (ADS)

Heidrich, Brenden J.

Nuclear power plants produce 20 percent of the electricity generated in the U.S. Nuclear generated electricity is increasingly valuable to a utility because it can be produced at a low marginal cost and it does not release any carbon dioxide. It can also be a hedge against uncertain fossil fuel prices. The construction of new nuclear power plants in the U.S. is cautiously moving forward, restrained by high capital costs. Since 1998, nuclear utilities have been increasing the power output of their reactors by implementing extended power up-rates. Power increases of up to 20 percent are allowed under this process. The equivalent of nine large power plants has been added via extended power up-rates. These up-rates require the replacement of large capital equipment and are often performed in concert with other plant life extension activities such as license renewals. This dissertation examines the effect of these extended power up-rates on the safety performance of U.S. boiling water reactors. Licensing event reports are submitted by the utilities to the Nuclear Regulatory Commission, the federal nuclear regulator, for a wide range of abnormal events. Two methods are used to examine the effect of extended power up-rates on the frequency of abnormal events at the reactors. The Crow/AMSAA model, a univariate technique is used to determine if the implementation of an extended power up-rate affects the rate of abnormal events. The method has a long history in the aerospace industry and in the military. At a 95-percent confidence level, the rate of events requiring the submission of a licensing event report decreases following the implementation of an extended power up-rate. It is hypothesized that the improvement in performance is tied to the equipment replacement and refurbishment that is performed as part of the up-rate process. The reactor performance is also analyzed using the proportional hazards model. This technique allows for the estimation of the effects of multiple independent variables on the event rate. Both the Cox and Weibull formulations were tested. The Cox formulation is more commonly used in survival analysis because of its flexibility. The best Cox model included fixed effects at the multi-reactor site level. The Weibull parametric formulation has the same base hazard rate as the Crow/AMSAA model. This theoretical connection was confirmed through a series of tests that demonstrated both models predicted the same base hazard rates. The Weibull formulation produced a model with most of the same statistically significant variables as the Cox model. The beneficial effect of extended power up-rates was predicted in the proportional hazards models as well as the Crow/AMSAA model. The Weibull model also indicated an effect that can be traced back to a plant’s construction. Performance was also found to improve in plants that had been divested from their original owners. This research developed a consistent evaluation toolkit for nuclear power plant performance using either a univariate method that allows for simple graphical evaluation at its heart or a more complex multivariate method that includes the effects of several independent variables with data that are available from public sources. Utilities or regulators with access to proprietary data may be able to expand upon this research with additional data that is not readily available to an academic researcher. Even without access to special data, the methods developed are valuable tools in evaluating and predicting nuclear power plant reliability performance.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy; Farmer, Mitchell; Corradini, Michael

The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized watermore » reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.« less

BOILING HOUSE, GROUND FLOOR. WAREHOUSE TO LEFT REAR, MASSECUITTE HEATERS ...

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

BOILING HOUSE, GROUND FLOOR. WAREHOUSE TO LEFT REAR, MASSECUITTE HEATERS ABOVE RIGHT, LOW GRADE CENTRIFUGALS BELOW. CRYSTALLIZER HOT WATER TANK TO REAR. VIEW FROM NORTHEAST - Lihue Plantation Company, Sugar Mill Building, Haleko Road, Lihue, Kauai County, HI

The causes and circumstances of drinking water incidents impact consumer behaviour: Comparison of a routine versus a natural disaster incident.

PubMed

Rundblad, Gabriella; Knapton, Olivia; Hunter, Paul R

2014-11-18

When public health is endangered, the general public can only protect themselves if timely messages are received and understood. Previous research has shown that the cause of threats to public health can affect risk perception and behaviours. This study compares compliance to public health advice and consumer behaviour during two "Boil Water" notices issued in the UK due to a routine incident versus a natural disaster incident. A postal questionnaire was sent to 1000 randomly selected households issued a routine "Boil Water" notice. Findings were then compared to a previous study that explored drinking water behaviour during a "Boil Water" notice issued after serious floods. Consumers affected by the routine incident showed a significant preference for official water company information, whereas consumers affected by the natural disaster preferred local information sources. Confusion over which notice was in place was found for both incidents. Non-compliance was significantly higher for the natural disaster (48.3%) than the routine incident (35.4%). For the routine incident, compliance with advice on drinking as well as preparing/cooking food and brushing teeth was positively associated with receiving advice from the local radio, while the opposite was true for those receiving advice from the water company/leaflet through the post; we suggest this may largely be due to confusion over needing boiled tap water for brushing teeth. No associations were found for demographic factors. We conclude that information dissemination plans should be tailored to the circumstances under which the advice is issued. Water companies should seek to educate the general public about water notices and which actions are safe and unsafe during which notice, as well as construct and disseminate clearer advice on brushing teeth and preparing/cooking food.

Detection of vapor nanobubbles by small angle neutron scattering (SANS)

NASA Astrophysics Data System (ADS)

Popov, Emilian; He, Lilin; Dominguez-Ontiveros, Elvis; Melnichenko, Yuri

2018-04-01

Experiments using boiling water on untreated (roughness 100-300 nm) metal surfaces using small-angle neutron scattering (SANS) show the appearance of structures that are 50-70 nm in size when boiling is present. The scattering signal disappears when the boiling ceases, and no change in the signal is detected at any surface temperature condition below saturation. This confirms that the signal is caused by vapor nanobubbles. Two boiling regimes are evaluated herein that differ by the degree of subcooling (3-10 °C). A polydisperse spherical model with a log-normal distribution fits the SANS data well. The size distribution indicates that a large number of nanobubbles exist on the surface during boiling, and some of them grow into large bubbles.

FILM-30: A Heat Transfer Properties Code for Water Coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

MARSHALL, THERON D.

2001-02-01

A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function ofmore » temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating.« less

Critical heat flux in subcooled flow boiling

NASA Astrophysics Data System (ADS)

Hall, David Douglas

The critical heat flux (CHF) phenomenon was investigated for water flow in tubes with particular emphasis on the development of methods for predicting CHF in the subcooled flow boiling regime. The Purdue University Boiling and Two-Phase Flow Laboratory (PU-BTPFL) CHF database for water flow in a uniformly heated tube was compiled from the world literature dating back to 1949 and represents the largest CHF database ever assembled with 32,544 data points from over 100 sources. The superiority of this database was proven via a detailed examination of previous databases. The PU-BTPFL CHF database is an invaluable tool for the development of CHF correlations and mechanistic models that are superior to existing ones developed with smaller, less comprehensive CHF databases. In response to the many inaccurate and inordinately complex correlations, two nondimensional, subcooled CHF correlations were formulated, containing only five adjustable constants and whose unique functional forms were determined without using a statistical analysis but rather using the parametric trends observed in less than 10% of the subcooled CHF data. The correlation based on inlet conditions (diameter, heated length, mass velocity, pressure, inlet quality) was by far the most accurate of all known subcooled CHF correlations, having mean absolute and root-mean-square (RMS) errors of 10.3% and 14.3%, respectively. The outlet (local) conditions correlation was the most accurate correlation based on local CHF conditions (diameter, mass velocity, pressure, outlet quality) and may be used with a nonuniform axial heat flux. Both correlations proved more accurate than a recent CHF look-up table commonly employed in nuclear reactor thermal hydraulic computer codes. An interfacial lift-off, subcooled CHF model was developed from a consideration of the instability of the vapor-liquid interface and the fraction of heat required for liquid-vapor conversion as opposed to that for bulk liquid heating. Severe vapor effusion in an upstream wetting front lifts the vapor-liquid interface off the surface, triggering CHF. Since the model is entirely based on physical observations, it has the potential to accurately predict CHF for other fluids and flow geometries which are beyond the conditions for which it was validated.

SASSYS pretest analysis of the THORS-SHRS experiments. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bordner, G.L.; Dunn, F.E.

The THORS Facility at ORNL was recently modified to allow the testing of two parallel 19-pin simulated fueled subassemblies under natural circulation conditions similar to those that might occur during a partial failure of the shutdown heat removal system (SHRS) of a liquid-metal fast breeder reactor. The planned experimental program included a series of tests at various inlet plenum temperatures to determine boiling threshold power levels and the power range for stable boiling during natural circulation operation. Pretest calculations were performed at ANL, which supplement those carried out at ORNL for the purposes of validating the SASSYS model in themore » natural circulation regime and of providing data which would be useful in planning the experiments.« less

Fundamental Boiling and RP-1 Freezing Experiments

NASA Technical Reports Server (NTRS)

Goode, Brian

2002-01-01

The prestart thermal conditioning of the hardware in LOX (liquid oxygen) systems involve heat transfer between LOX and metal where boiling plays a large role. Information is easily found on nucleate boiling, maximum heat flux, minimum heat flux and film boiling for common fluids like water. After looking at these standard correlations it was felt more data was needed for the cool down side transition boiling for the LN2 and LOX. In particular interest is the film boiling values, the temperature at which transition begins and the slope as peak heat flux is approached. The ultimate goal is an array of boiling heat transfer coefficient as a function of surface temperature which can be used in the chilldown model of the feed system, engine and bleed system for X-34. The first experiment consisted of an actual MC-1 LOX Impeller which had been machined backwards, that was instrumented with 17 surface thermocouples and submerged in liquid nitrogen. The thermocouples were installed on metal thicknesses varying from the thin inducer to the thick hub.

Does pan diameter influence carbon monoxide levels during heating of water to boiling point with a camping stove?

PubMed

Leigh-Smith, Simon; Stevenson, Richard; Watt, Martin; Watt, Ian; McFadyen, Angus; Grant, Stan

2004-01-01

To determine whether pan diameter influences carbon monoxide (CO) concentration during heating of water to boiling point with a camping stove. The hypothesis was that increasing pan diameter increases CO concentration because of greater flame dispersal and a larger flame. This was a randomized, prospective study. A Coleman Dual Fuel 533 stove was used to heat pans of water to boiling point, with CO concentration monitored every 30 seconds for 5 minutes. The stove was inside a partially ventilated 200-L cardboard box model that was inside an environmental chamber at -6 degrees C. Water temperature, water volume, and flame characteristics were all standardized. Ten trials were performed for each of 2 pan diameters (base diameters of 165 mm [small] and 220 mm [large]). There was a significant difference (P = .002) between the pans for CO levels at each measurement interval from 60 seconds onward. These differences were markedly larger after 90 seconds, with a mean difference of 185 ppm (95% CI 115, 276 ppm) for all the results from 120 seconds onwards. This study has shown that there is significantly higher CO production with a large-diameter pan compared with a small-diameter pan. These findings were evident by using a camping stove to heat water to boiling point when a maximum blue flame was present throughout. Thus, in enclosed environments it is recommended that small-diameter pans be used in an attempt to prevent high CO levels.

Removal of 16 pesticide residues from strawberries by washing with tap and ozone water, ultrasonic cleaning and boiling.

PubMed

Lozowicka, Bozena; Jankowska, Magdalena; Hrynko, Izabela; Kaczynski, Piotr

2016-01-01

The effects of washing with tap and ozone water, ultrasonic cleaning and boiling on 16 pesticide (ten fungicides and six insecticides) residue levels in raw strawberries were investigated at different processing times (1, 2 and 5 min). An analysis of these pesticides was conducted using gas chromatography with nitrogen-phosphorous and electron capture detection (GC-NPD/ECD). The processing factor (PF) for each pesticide in each processing technique was determined. Washing with ozonated water was demonstrated to be more effective (reduction from 36.1 to 75.1 %) than washing with tap water (reduction from 19.8 to 68.1 %). Boiling decreased the residues of the most compounds, with reductions ranging from 42.8 to 92.9 %. Ultrasonic cleaning lowered residues for all analysed pesticides with removal of up to 91.2 %. The data indicated that ultrasonic cleaning and boiling were the most effective treatments for the reduction of 16 pesticide residues in raw strawberries, resulting in a lower health risk exposure. Calculated PFs for alpha-cypermethrin were used to perform an acute risk assessment of dietary exposure. To investigate the relationship between the levels of 16 pesticides in strawberry samples and their physicochemical properties, a principal component analysis (PCA) was performed. Graphical abstract ᅟ.