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Sample records for abwr advanced boiling

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

    Energy Technology Data Exchange (ETDEWEB)

    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 readily 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.

  2. Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

    International Nuclear Information System (INIS)

    The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

  3. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  4. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  5. Development of Innovative Construction Technologies for ABWRs

    International Nuclear Information System (INIS)

    This paper describes an effort by Tokyo Electric Power Company (TEPCO) to shorten the construction time in a drastic manner for the Advanced Boiling Water Reactors (ABWR), thereby aiming at reducing construction costs. First an outline of the actual construction records for the five BWR Units and the two ABWR Units at the Kashiwazaki-Kariwa site is introduced along with the construction methods employed for these units. There is a continued trend in the reduction of construction time from Unit 1 to 7 owing to a number of improvements made to these units, and in particular it is noteworthy that the drastic reduction was accomplished due to the change in reactor type from BWR to ABWR. Explained next is an on-going effort for the next ABWR and the following next generation of ABWRs to further shorten construction time. In this effort an emphasis is laid on the development of innovative construction methods by the adoption of steel plate/concrete composite structures (SC Structure), and the application of those structures even to containment vessel (so-called SCCV). This work is being conducted as part of TEPCO's research and development for the next generation ABWRs (ABWR II). Presently it is expected that the construction time from bedrock inspection to fuel loading could be shortened in a stepwise fashion to 38 and 31.5 months for the next ABWR and finally to 23 months for the next generation ABWR II, thus enabling a greatly reduced power generating cost and enhanced safety during construction. In addition to the above effort, a preliminary study has been performed concerning the application of base isolation systems to ABWR plants. From this study which is aimed at standardization, lowered cost and mitigation of engineering work, it was found that base isolation system effectively reduces seismic response and is instrumental in achieving the above-mentioned objectives. Therefore it can be said that the use of this system gives more freedom in selecting sites

  6. Advanced plant engineering and construction of Japanese ABWRs

    International Nuclear Information System (INIS)

    Remarkable improvement has been made in recent nuclear power plant design and construction in Japan. These many improved engineering technologies has been made a good use in the lately commercial operated two world's first 1,356MWe ABW's (Advanced Boiling Water Reactors), and made a great contribution to the smooth progress and the completion of a highly reliable plant construction. Especially, two engineering technologies, (1), three-dimensional computer aided design system through engineering data-base, and (2), large scale modularising construction method, have been successfully applied as the integrated engineering technologies of the plant construction. And two integrated reviews, 'integrated design review, confirmation of new and changed design and prevention of failure recurrence' in the design stage, and 'constructing plant review' at the site, have been widely and systematically conducted as a link in the chain of steady reliability improvement activities. These advanced and/or continuous and steady technologies are one of most important factors for high reliability through the entire lifetime of a nuclear plant, including planning, design, construction, operation and maintenance stages. (author)

  7. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  8. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE-ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  9. Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

    International Nuclear Information System (INIS)

    This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff's review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff's review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    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, 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

  11. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    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, 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

  12. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    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, 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.

  13. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    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, 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.

  14. Development of Advanced Concept for Shortening Construction Period of ABWR Plant

    International Nuclear Information System (INIS)

    Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

  15. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Science.gov (United States)

    2010-01-01

    ...) of 10 CFR 50.34—Post-Accident Sampling for Boron, Chloride, and Dissolved Gases; and 3. Paragraph (f... design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that... design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10...

  16. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  17. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  18. Development of advanced concept for shortening construction period of ABWR plant (Part 3)

    International Nuclear Information System (INIS)

    The reinforced concrete containment vessel (RCCV) is the most critical part in construction of an ABWR plant. Use of steel plate reinforced concrete (SC) and a large modular construction method are effective in shortening the construction period (Ijichi et al., 2004)1). This research is aimed at a remarkable shortening of the construction period of an ABWR plant (period from 1st concrete placement to Fuel/Loading is less than 22 months). Conceptual design of a steel plate reinforced concrete containment vessel (SCCV) using an SC structure is carried out and structural experiments are conducted. It is thus confirmed that SCCV shows outstanding structural performance, compared with RCCV. This paper outlines the study results. (authors)

  19. Modeling pressure suppression pool hydrodynamics in the ABWR containment

    International Nuclear Information System (INIS)

    Highlights: → Developed mechanistic model for prediction of suppression pool hydrodynamics parameters. → Vent clearance time, pool swell height, and bubble breakthrough elevation predicted within 10% of the experimental data. → Performed assessment of pressure suppression pool hydrodynamics in ABWR. → Reasonable agreement obtained between the model predictions and the licensing analyses. - Abstract: This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.

  20. 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

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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 constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  1. A compilation of boiling water reactor operational experience for the United Kingdom's Office for Nuclear Regulations Advanced Boiling Water Reactor generic design assessment.

    Energy Technology Data Exchange (ETDEWEB)

    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 Kingdoms Health and Safety Executive Office for Nuclear Regulations (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 constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

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

    Energy Technology Data Exchange (ETDEWEB)

    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 Population 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.

  3. Safety design philosophy of the ABWR for the next generation LWRs

    International Nuclear Information System (INIS)

    The paper presents safety design philosophy of the advanced boiling water reactor (ABWR) to be reflected in developing the next generation light water reactors (LWRs). The basic policy of the ABWR safety design was to improve safety and reduce cost simultaneously by reflecting lessons learned of precursors, incidents and accidents that were beyond the design basis such as the Three Mile Island Unit 2 (TMI 2) accident. The ABWR is a fully active safety plant. The ABWR enhanced redundancy and diversity of active safety systems using probabilistic safety assessment (PSA) insights. It adopted a complete three division active emergency core cooling system (ECCS) and attained a very low core damage frequency (CDF) value of less than 10-7/ry for internal events. Only very small residual risks, if any, rather exist in external events such as an extremely large earthquake beyond the design basis. This is because external events can constitute a common cause that disables all the redundant active safety systems. Therefore, it is useless to add one more ECCS train and make a four division active ECCS for external events. Nowadays, however, fully passive safety LWRs are already established. Incorporating some of these passive safety systems we can also establish the next generation LWRs that are truly strong against external events. We can establish a plant that can survive a giant earthquake at least three days without AC power source, SA proof safety design that enables no containment failure and no evacuation to eliminate the residual risks. The same basic policy as the ABWR to improve safety and reduce cost simultaneously is again effective for the next generation LWRs. (author)

  4. Defence-in-depth concept for the EU-ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Hiroshi; Fuchs, Steffen; Takada, Toshiaki; Kataoka, Kazuyoshi [Toshiba International Limited (Japan)

    2013-07-01

    The current defence-in-depth (DiD) concept has been established by the Reactor Harmonization Working Group (RHWG) of Western European Nuclear Regulators Association (WENRA). Principally the DiD concept was already part of the very early power reactor designs. However, additional considerations have been done in order to take plant conditions into account which are beyond the original design basis. The most recent advancements have been done based on major lessons learned from the Fukushima Dai-Ichi accident. Especially for new nuclear reactors it has to be demonstrated that DiD aspects have been considered in their design. Currently Toshiba is adapting its Advanced Boiling Water Reactor (ABWR) for the European market, at first in Finland. This presentation aims to describe how the new DiD concept has been applied to achieve the safety goals for a modern reactor type and to ensure a design that can be licensed in Western Europe. (orig.)

  5. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  6. Improvements in boiling water reactor designs and safety

    International Nuclear Information System (INIS)

    The advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are discussed in this paper. They include: design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 last stage buckets; and advanced radwaste technology

  7. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. 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 permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US 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 typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  8. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    International Nuclear Information System (INIS)

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. 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 permanent conversion and long-term diversion resistance to this material. The NAS study ''Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US 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 typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs

  9. Study of Pu consumption in Advanced Light Water Reactors

    International Nuclear Information System (INIS)

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

  10. Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

    International Nuclear Information System (INIS)

    An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

  11. Boils

    Science.gov (United States)

    ... the boil is very bad or comes back. Antibacterial soaps and creams cannot help much once a boil ... following may help prevent the spread of infection: Antibacterial soaps Antiseptic (germ-killing) washes Keeping clean (such as ...

  12. Boiling heat transfer modern developments and advances

    CERN Document Server

    Lahey, Jr, RT

    2013-01-01

    This volume covers the modern developments in boiling heat transfer and two-phase flow, and is intended to provide industrial, government and academic researchers with state-of-the-art research findings in the area of multiphase flow and heat transfer technology. Special attention is given to technology transfer, indicating how recent significant results may be used for practical applications. The chapters give detailed technical material that will be useful to engineers and scientists who work in the field of multiphase flow and heat transfer. The authors of all chapters are members of the

  13. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  14. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  15. World's first ABWR start-up test analysis with 3-D transient computational code

    International Nuclear Information System (INIS)

    The Kashiwazaki-Kariwa Nuclear Power Station Unit 6, the world's first Advanced BWR (ABWR), began commercial operation from November 1996 following one year of start-up tests. A large number of variables which may be used to validate the advanced design features were obtained from transient tests. These test data are now being used for the qualification of TRACG, a BWR 3-D transient analysis code. Calculated results show that TRACG is fully capable of accurately predicting ABWR transient response and will be useful for application to future plant designs

  16. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  17. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-02-22

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  18. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-03-09

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling... October 14, 2009, (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available...

  19. The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

    International Nuclear Information System (INIS)

    The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

  20. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  1. Assessment of Analytical Prediction of JNES Seismic Wall Pressure Data for ABWR Model Structures

    International Nuclear Information System (INIS)

    Prior to the establishment of the Japan Nuclear Energy Safety Organization (JNES), the Nuclear Power Engineering Corporation (NUPEC) of Japan conducted a series of large-scale field tests for the Ministry of Economy, Trade and Industry (METI) of Japan to address various aspects of the soil-structure interaction (SSI) effect on the seismic response of nuclear power plant (NPP) structures. The experimental studies used several scaled models of advanced boiling water reactor (ABWR) structures, which were constructed at field sites typical of an actual NPP site. As part of the US-Japan collaboration effort on the soil-structure interaction (SSI) phenomenon, the U.S. Nuclear Regulatory Commission (NRC) and Brookhaven National Laboratory (BNL) performed a study to correlate the recorded earthquake induced wall pressures using commercial programs. The purpose of this study is to assess the adequacy and performance of the analytical SSI methods in predicting recorded data, and to determine the effect and breadth of the soil uncertainty on the seismic response computations to capture seismic induced passive soil pressures. The commercial programs used for the correlation study are SASSI and LS-DYNA. SASSI uses the sub-structuring method, which treats SSI responses by superimposing the finite element model of the structure with the analytical solution of wave propagation in the half-space. LS-DYNA is an explicit finite element program, and is only appropriate to treating problems which can be defined with finite boundaries. Therefore, the half-space SSI problem is approximately modeled using LS-DYNA with explicit finite elements for both structure and the surrounding soil to the extent that the scattered waves resulting from the structural vibration will not be reflected at the soil boundaries. This is further ensured by placing the Lysmer transmitting elements at the soil mesh boundaries to absorb the outgoing waves. This paper presents an overview of the NRC

  2. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

    Ehud Greenspan; Jasmina Vujic; Francesco Ganda; Arias, Francisco J.

    2012-01-01

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR) proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorber...

  3. A Matrix Method of Analyzing the Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit%先进型沸水堆核电机组热经济性矩阵分析方法

    Institute of Scientific and Technical Information of China (English)

    冉鹏; 李庚生; 廖丹; 朱伟平

    2010-01-01

    根据先进型沸水堆(advance boiling water reactor,ABWR)核电机组热力系统的结构特点,基于热力系统等效热降分析方法和矩阵方法,确定其主、辅系统的划分原则以及辅助汽水成分划分原则,对先进型沸水堆各种汽水成分进行归并处理,构建表达规则的先进型沸水堆核电机组汽水分布方程填写规则,推导出适合先进型沸水堆核电机组热力系统热经济性分析的通用矩阵方法,并给出该类型核电机组辅助汽水成分对热经济性影响的表达方式.该矩阵全面反映了先进犁沸水堆核电机组热力系统主系统和各种辅助系统对机组热经济性的影响状况,每个子矩阵物理意义明确、规律性强,可使先进型沸水堆核电机组热力系统的整体计算和局部分析变得清晰、简单,适合于计算机程序化,并通过实例对该方法进行了验证.

  4. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    International Nuclear Information System (INIS)

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval

  5. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  6. Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface Tracking Method

    Science.gov (United States)

    Yoshida, Hiroyuki; Takase, Kazuyuki

    Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low.

  7. Accumulation of operator workload data by using A-BWR training simulator

    International Nuclear Information System (INIS)

    Human-machine interface (HMI) of A-BWR has been developed in order to improve operational safety, reliability and to reduce workload. A-BWR HMI is fully computerized. JNES (Japan Nuclear Energy Safety Organization) and BTC (BWR Operator Training Center Corporation) have accumulated the operator workload quantitative data, related to the observation and operation at typical transient conditions, in order to evaluate the difference of operational workloads between A-BWR HMI and conventional type HMI. The workload evaluation shows the following results: - The workload density (observation and operation frequencies per unit time) of ABWR just after the plant trip is less than that of BWR-5. - At stable conditions after the transient, the workload density of ABWR becomes higher comparing that of BWR-5. A-BWR alarm system may increase the workload density caused by alarm multi-layer structure, because an operator has to use the flat and/or the CRT display to pursue every alarm. The analysis of shift team training at BTC shows that total workload is reduced at ABWR but alarm confirmation work still remains as burden. These results show some modifications might be needed for future HMI. To grasp the tendency of operator workload difference by the control panel type difference, the operator workload quantitative data have accumulated using ABWR type simulator and conventional type simulator at the same typical transient condition. These data were arranged operation frequency data, number of alarm generating data and number of switching CRT pictures data according to plant behaviour. The ABWR type HMI's characteristic have become clear by these operator workload data and the team characteristic evaluation data which BTC evaluated comparing the team performance difference of HMI type

  8. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  9. Advanced modeling of the size poly-dispersion of boiling flows

    International Nuclear Information System (INIS)

    Full text of publication follows: This work has been performed within the Institut de Radioprotection et de Surete Nucleaire that leads research programs concerning safety analysis of nuclear power plants. During a LOCA (Loss Of Coolant Accident), in-vessel pressure decreases and temperature increases, leading to the onset of nucleate boiling. The present study focuses on the numerical simulation of the local topology of the boiling flow. There is experimental evidence of a local and statistical large spectra of possible bubble sizes. The relative importance of the correct description of this poly-dispersion in size is due to the dependency of (i) main hydrodynamic forces, like lift, as well as of (ii) transfer area with respect to the individual bubble size. We study the corresponding CFD model in the framework of an ensemble averaged description of the dispersed two-phase flow. The transport equations of the main statistical moment densities of the population size distribution are derived and models for the mass, momentum and heat transfers at the bubble scale as well as for bubble coalescence are achieved. This model introduced within NEPTUNE-CFD code of the NEPTUNE thermal-hydraulic platform, a joint project of CEA, EDF, IRSN and AREVA, has been tested on boiling flows obtained on the DEBORA facility of the CEA at Grenoble. These numerical simulations provide a validation and attest the impact of the proposed model. (authors)

  10. The development and assessment of TRACE/PARCS model for Lungmen ABWR

    International Nuclear Information System (INIS)

    Highlights: • The development of a TRACE/PARCS model of Lungmen (ABWR). • Thermal-hydraulics model of TRACE is coupling with 3-D neutronics model of PARCS. • The 3-D geometry vessel component of TRACE is used in this study. • The parameters responses of Lungmen TRACE/PARCS model are consistent with FSAR data. - Abstract: This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass and pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses

  11. ABWR, an option for the electric generation in Mexico

    International Nuclear Information System (INIS)

    The A BWR reactor (Advanced Boiling Water Reactor) it was developed in a project group among the Company TEPCO, (Tokyo Electric Power Company), Hitachi, Toshiba and General Electric. The A BWR is the first nuclear reactor of the type BWR third generation that entered in commercial operation in the 90 Th decade. One of those main characteristics of the A BWR are that the system of external recirculation has been eliminated, that is to say, the pumps and external recirculation pipes have been replaced by 10 internal recirculation pumps mounted in the inferior part of the pressure vessel, for that external recirculation systems neither the use of jet pumps are not needed. Another important characteristic of the A BWR is the simplification of the activation of the safety systems. The simplifications in the design of the A BWR and the use of new technologies have reduced the quantity of equipment and the time of construction compared with the previous designs of BWRs. The construction project for the A BWR consists of a period of construction from 48 to 54 months, measured since that the first concrete structure is placed until that it enters in commercial operation, in accordance with the documents liberated by G E. (Author)

  12. ABWR, an option for the electric generation in Mexico; ABWR, una opcion para la generacion electrica en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Gomez T, A.M.; Ramirez S, J.R.; Xolocostli M, J.V. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: amgt@nuclear.inin.mx

    2005-07-01

    The A BWR reactor (Advanced Boiling Water Reactor) it was developed in a project group among the Company TEPCO, (Tokyo Electric Power Company), Hitachi, Toshiba and General Electric. The A BWR is the first nuclear reactor of the type BWR third generation that entered in commercial operation in the 90 Th decade. One of those main characteristics of the A BWR are that the system of external recirculation has been eliminated, that is to say, the pumps and external recirculation pipes have been replaced by 10 internal recirculation pumps mounted in the inferior part of the pressure vessel, for that external recirculation systems neither the use of jet pumps are not needed. Another important characteristic of the A BWR is the simplification of the activation of the safety systems. The simplifications in the design of the A BWR and the use of new technologies have reduced the quantity of equipment and the time of construction compared with the previous designs of BWRs. The construction project for the A BWR consists of a period of construction from 48 to 54 months, measured since that the first concrete structure is placed until that it enters in commercial operation, in accordance with the documents liberated by G E. (Author)

  13. Boils (Furunculosis)

    Science.gov (United States)

    ... boil starts to drain, wash the area with antibacterial soap and apply some triple antibiotic ointment and a ... avoid spreading the infection to others. Use an antibacterial soap on boil-prone areas when showering, and dry ...

  14. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  15. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  16. An advanced frequency-domain code for boiling water reactor (BWR) stability analysis and design

    International Nuclear Information System (INIS)

    The two-phase flow instability is of interest for the design and operation of many industrial systems such as boiling water reactors (BWRs), chemical reactors, and steam generators. In case of BWRs, the flow instabilities are coupled to the power instabilities via neutronic-thermal hydraulic feedbacks. Since these instabilities produce also local pressure oscillations, the coolant flashing plays a very important role at low pressure. Many frequency-domain codes have been used for two-phase flow stability analysis of thermal hydraulic industrial systems with particular emphasis to BWRs. Some were ignoring the effect of the local pressure, or the effect of 3D power oscillations, and many were not able to deal with the neutronics-thermal hydraulics problems considering the entire core and all its fuel assemblies. The new frequency domain tool uses the best available nuclear, thermal hydraulic, algebraic and control theory methods for simulating BWRs and analyzing their stability in either off-line or on-line fashion. The novel code takes all necessary information from plant files via an interface, solves and integrates, for all reactor fuel assemblies divided into a number of segments, the thermal-hydraulic non-homogenous non-equilibrium coupled linear differential equations, and solves the 3D, two-energy-group diffusion equations for the entire core (with spatial expansion of the neutron fluxes in Legendre polynomials).It is important to note that the neutronics equations written in terms of flux harmonics for a discretized system (nodal-modal equations) generate a set of large sparse matrices. The eigenvalue problem associated to the discretized core statics equations is solved by the implementation of the implicit restarted Arnoldi method (IRAM) with implicit shifted QR mechanism. The results of the steady state are then used for the calculation of the local transfer functions and system transfer matrices. The later are large-dense and complex matrices, (their size

  17. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling

  18. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  19. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  20. Oscillate Boiling

    CERN Document Server

    Li, Fenfang; Nguyen, Dang Minh; Ohl, Claus-Dieter

    2016-01-01

    We report about an intriguing boiling regime occurring for small heaters embedded on the boundary in subcooled water. The microheater is realized by focusing a continuous wave laser beam to about $10\\,\\mu$m in diameter onto a 165\\,nm-thick layer of gold, which is submerged in water. After an initial vaporous explosion a single bubble oscillates continuously and repeatably at several $100\\,$kHz. The microbubble's oscillations are accompanied with bubble pinch-off leading to a stream of gaseous bubbles into the subcooled water. The self-driven bubble oscillation is explained with a thermally kicked oscillator caused by the non-spherical collapses and by surface pinning. Additionally, Marangoni stresses induce a recirculating streaming flow which transports cold liquid towards the microheater reducing diffusion of heat along the substrate and therefore stabilizing the phenomenon to many million cycles. We speculate that this oscillate boiling regime may allow to overcome the heat transfer thresholds observed dur...

  1. Basic research and industrialization of CANDU advanced fuel - Effect of transverse convex curvature on boiling heat transfer and ONB point of nucleate fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Chun; Lee, Young; Lee, Sung Hong [Pusan National University, Pusan (Korea)

    2000-04-01

    Recently, the effect of convex curvature on heat transfer should not be ignored when the radius of curvature tends to be small and/or associated with high heat transfer rate cases. Both analytical and experimental studies were performed to prove the effect of transverse convex curvature on the boiling heat transfer in concentric annuli flows. The effect of the transverse convex surface curvature on ONB are studied analytically in the case of reactor and evaporator. It is seen that the inner wall heat flux depends on R/sub i/, Rc, Re, Pr, {alpha}, and the {theta} of working fluid. An experimental study on the incipience of nucleate boiling is performed as a verification ad extension of previous analyses. Through flow visualization, the results show that the most dominant parameter to affect the heat flux at ONB is found to be the surface curvature. The heat flux data at ONB increases with the Re and the subcooling, and the effect of subcooling on ONB becomes smaller with decreasing Re. The heat flux at ONB increases rapidly as increase in {alpha} due to higher convective motion of bulk flow. Comparison between both results are accomplished with respect to the relative enhancement due to the convex curvature. The relative heat transfer enhancement ratio shows a good agreement between theory and experiment qualitatively and quantitatively. In conclusion, the obtained results suggest that the effect transverse convex curvature appears significantly in the boiling heat transfer. Therefore, it can be clearly expected that the effect should be more strong at the case of critical heat flux condition which is the most important design goal of the advanced nuclear fuel rods. 30 refs., 78 figs. (Author)

  2. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  3. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  4. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  5. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  6. MELCOR 1.8.2 calculations of selected sequences for the ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.

    1994-07-01

    This report summarizes the results from MELCOR calculations of severe accident sequences in the ABWR and presents comparisons with MAAP calculations for the same sequences. MELCOR was run for two low-pressure and three high-pressure sequences to identify the materials which enter containment and are available for release to the environment (source terms), to study the potential effects of core-concrete interaction, and to obtain event timings during each sequence; the source terms include fission products and other materials such as those generated by core-concrete interactions. Sensitivity studies were done on the impact of assuming limestone rather than basaltic concrete and on the effect of quenching core debris in the cavity compared to having hot, unquenched debris present.

  7. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

  8. Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

    International Nuclear Information System (INIS)

    This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A ampersand 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met

  9. Secondary pool boiling effects

    Science.gov (United States)

    Kruse, C.; Tsubaki, A.; Zuhlke, C.; Anderson, T.; Alexander, D.; Gogos, G.; Ndao, S.

    2016-02-01

    A pool boiling phenomenon referred to as secondary boiling effects is discussed. Based on the experimental trends, a mechanism is proposed that identifies the parameters that lead to this phenomenon. Secondary boiling effects refer to a distinct decrease in the wall superheat temperature near the critical heat flux due to a significant increase in the heat transfer coefficient. Recent pool boiling heat transfer experiments using femtosecond laser processed Inconel, stainless steel, and copper multiscale surfaces consistently displayed secondary boiling effects, which were found to be a result of both temperature drop along the microstructures and nucleation characteristic length scales. The temperature drop is a function of microstructure height and thermal conductivity. An increased microstructure height and a decreased thermal conductivity result in a significant temperature drop along the microstructures. This temperature drop becomes more pronounced at higher heat fluxes and along with the right nucleation characteristic length scales results in a change of the boiling dynamics. Nucleation spreads from the bottom of the microstructure valleys to the top of the microstructures, resulting in a decreased surface superheat with an increasing heat flux. This decrease in the wall superheat at higher heat fluxes is reflected by a "hook back" of the traditional boiling curve and is thus referred to as secondary boiling effects. In addition, a boiling hysteresis during increasing and decreasing heat flux develops due to the secondary boiling effects. This hysteresis further validates the existence of secondary boiling effects.

  10. Evaporation, Boiling and Bubbles

    Science.gov (United States)

    Goodwin, Alan

    2012-01-01

    Evaporation and boiling are both terms applied to the change of a liquid to the vapour/gaseous state. This article argues that it is the formation of bubbles of vapour within the liquid that most clearly differentiates boiling from evaporation although only a minority of chemistry textbooks seems to mention bubble formation in this context. The…

  11. Aspects of subcooled boiling

    Energy Technology Data Exchange (ETDEWEB)

    Bankoff, S.G. [Northwestern Univ., Evanston, IL (United States)

    1997-12-31

    Subcooled boiling boiling refers to boiling from a solid surface where the bulk liquid temperature is below the saturation temperature (subcooled). Two classes are considered: (1) nucleate boiling, where, for large subcoolings, individual bubbles grow and collapse while remaining attached to the solid wall, and (2) film boiling, where a continuous vapor film separates the solid from the bulk liquid. One mechanism by which subcooled nucleate boiling results in very large surface heat transfer coefficient is thought to be latent heat transport within the bubble, resulting from simultaneous evaporation from a thin residual liquid layer at the bubble base, and condensation at the polar bubble cap. Another is the increased liquid microconvection around the oscillating bubble. Two related problems have been attacked. One is the rupture of a thin liquid film subject to attractive and repulsive dispersion forces, leading to the formation of mesoscopic drops, which then coalesce and evaporate. Another is the liquid motion in the vicinity of an oscillating contact line, where the bubble wall is idealized as a wedge of constant angle sliding on the solid wall. The subcooled film boiling problem has been attacked by deriving a general long-range nonlinear evolution equation for the local thickness of the vapor layer. Linear and weakly-nonlinear stability results have been obtained. A number of other related problems have been attacked.

  12. A review on boiling heat transfer enhancement with nanofluids.

    Science.gov (United States)

    Barber, Jacqueline; Brutin, David; Tadrist, Lounes

    2011-04-04

    There has been increasing interest of late in nanofluid boiling and its use in heat transfer enhancement. This article covers recent advances in the last decade by researchers in both pool boiling and convective boiling applications, with nanofluids as the working fluid. The available data in the literature is reviewed in terms of enhancements, and degradations in the nucleate boiling heat transfer and critical heat flux. Conflicting data have been presented in the literature on the effect that nanofluids have on the boiling heat-transfer coefficient; however, almost all researchers have noted an enhancement in the critical heat flux during nanofluid boiling. Several researchers have observed nanoparticle deposition at the heater surface, which they have related back to the critical heat flux enhancement.

  13. Dry patch formed boiling and burnout in potassium pool boiling

    International Nuclear Information System (INIS)

    Experimental results are presented on dry patch formed boiling and burnout in saturated potassium pool boiling on a horizontal plane heater for system pressures from 30 to 760 torr and liquid levels from 5 to 50 mm. The dry patch formation occurs in the intermittent boiling which is often encountered when liquid alkali metals are used under relatively low pressure conditions. Burnout is caused from both continuous nucleate and dry patch formed boiling. The burnout heat flux together with nucleate boiling heat transfer coefficients are empirically correlated with system pressures. A model is also proposed to predict the minimum heat flux to form the dry patch. (author)

  14. Advances in the development and validation of CFD-BWR, a two-phase computational fluid dynamics model for the simulation of flow and heat transfer in boiling water reactors

    International Nuclear Information System (INIS)

    This paper presents recent advances in the validation of an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel bundles. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. We have described the model development strategy that has been adopted by the development team for the prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow topology maps and flow topology specific phenomenological models. The paper reviews the key boiling phenomenological models and focuses on recent results of experiment analyses for the validation of two-phase BWR phenomena models including cladding-to-coolant heat transfer and Critical Heat Flux experiments and the BWR Full-size Assembly Boiling Test (BFBT). The two-phase flow models implemented in the CFD-BWR code can be grouped into three broad categories: models describing the vapor generation at the heated cladding surface, models describing the interactions between the vapor and the liquid coolant, and models describing the heat transfer between the fuel pin and the two-phase coolant. These models have been described and will be briefly reviewed. The boiling model used in the second generation of the CFD-BWR code includes a local flow topology map which allows the cell-by-cell selection of the local flow topology. Local flow topologies can range from a bubbly flow topology where the continuous phase is liquid, to a transition flow topology, to a droplet flow topology where the continuous phase is vapor, depending primarily on the local void fraction. The models describing the cladding-to-coolant heat transfer and the interplay between these models and the local flow topology are important in Critical Heat Flux (CHF) analyses, and will

  15. ABWR start-up test analysis using BWR core simulator with three-dimensional direct response matrix method

    International Nuclear Information System (INIS)

    The ABWR start-up test analysis has been done with the BWR core simulator using the three--dimensional direct response matrix (3D-DRM) method. The Monte Carlo code VMONT made the sub-response matrices for the 3D-DRM method. Each boundary surface was subdivided by 4 x 4 for transverse segments, by 4 for angular segments and by 4 for axial zones in a node. For the calculation speedup, the 3D-DRM code used the divided sub-response matrices data set. The code used the MPI and OpenMP for the parallelized method. The median value is set as the average critical eigenvalues. The changes from the maximum value to the minimum value are 0.34 %Δk with the spectral history method and 0.40 %Δk without it, and the respective standard deviations were 0.12 % and 0.14 %. Using the spectral history method decreased the variation by 0.06 %Δk. The root mean square differences of the axial power distribution were about 6 % between the analysis results and the plant data. Using the currents which converged in the previous exposure step reduced the number of iterations when the CR pattern changed only slightly. The averaged calculation time for each exposure step was about 5 hours on 12 PC Linux cluster servers with Core 2 Quad 3 GHz. (authors)

  16. 3RIP trip startup test simulation of TRACE/PARCS model for Lungmen ABWR under different power and flow conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Shu-Ming; Wang, Jong-Rong; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., Hsin Chu, Taiwan (China). Inst. of Nuclear Engineering and Science; Lin, Hao-Tzu [Atomic Energy Council, Taoyuan City, Taiwan (China). Inst. of Nuclear Energy Research

    2015-12-15

    Lungmen nuclear power plant project started long time ago, it is not yet commercially operated but Taiwan Power Company has already prepared for its startup tests. 3RIP trip startup test is one of them. Three of the 10 RIPs will be manually tripped in the test. Response of the plant for this transient will be watched and recorded to check if the test criteria are satisfied. This paper is a result of code simulation of 3RIP trip startup test of Lungmen ABWR nuclear power plant. Thermal hydraulic code TRACE coupled with neutronics code PARCS were used to build the simulation model of Lungmen nuclear power plant. Startup tests under different plant power and flow conditions are considered in this research. A sensitivity study on the impact of different pump moment of inertia has been performed. Simulation results with TRACE and PARCS shows that the acceptance criteria for this startup test can be satisfied and the impact of different pump inertia is little.

  17. Geysering in boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Aritomi, Masanori; Takemoto, Takatoshi [Tokyo Institute of Technology, Tokyo (Japan); Chiang, Jing-Hsien [Japan NUS Corp. Ltd., Toyko (Japan)] [and others

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  18. Boiling incipience and convective boiling of neon and nitrogen

    Science.gov (United States)

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

    1977-01-01

    Forced convection and subcooled boiling heat transfer data for liquid nitrogen and liquid neon were obtained in support of a design study for a 30 tesla cryomagnet cooled by forced convection of liquid neon. This design precludes nucleate boiling in the flow channels as they are too small to handle vapor flow. Consequently, it was necessary to determine boiling incipience under the operating conditions of the magnet system. The cryogen data obtained over a range of system pressures, fluid flow rates, and applied heat fluxes were used to develop correlations for predicting boiling incipience and convective boiling heat transfer coefficients in uniformly heated flow channels. The accuracy of the correlating equations was then evaluated. A technique was also developed to calculate the position of boiling incipience in a uniformly heated flow channel. Comparisons made with the experimental data showed a prediction accuracy of plus or minus 15 percent

  19. Steady State Vapor Bubble in Pool Boiling.

    Science.gov (United States)

    Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C; Maroo, Shalabh C

    2016-01-01

    Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics. PMID:26837464

  20. Steady State Vapor Bubble in Pool Boiling

    Science.gov (United States)

    Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C.; Maroo, Shalabh C.

    2016-02-01

    Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics.

  1. Investigation on the design of human-system interface for advanced nuclear plant control room

    International Nuclear Information System (INIS)

    The Lungmen Nuclear Power Project (LMNPP), under construction in Taiwan, consists of two GE Advanced Boiling Water Reactor (ABWR) units, each with 1350 MW electrical output. Major Human-System Interfaces (HSIs) of LMNPP are different from traditional ones. Video display units (VDUs) are the main human-system interface for operators to manipulate and to know the status of the equipment and plant information. Based upon NUREG-0711, the applicable human factors engineering (HFE) guideline in the design of HSIs has been adopted. An important aspect of the Lungmen HFE program has been the direct involvement of the end user, Taiwan Power Company (TPC), throughout the design development and implementation to ensure not only that the process for the design is compliant with the HFE principles, but also that the necessary displays, control, and alarms are provided to support the identified personnel tasks. This paper reviews the applicable HFE principles and verification and validation (V and V) processes in the design of HSIs for the advanced LMNPP. This paper also presents three investigated topics of the LMNPP HSI design development and implementation process. From the perspective of licensing concern and experience feedback, the focus of this paper is on the topics of validation with simulator, alarm auto reset, and VDU operational configuration strategy. The objectives of investigating the latter topic were to ensure the VDU operational configuration strategy, after appropriate V and V, achieves its goals of reinforcing operation discipline and distributing operator crew task assignments and workload during typical operations, and to confirm the need for an intensive training program that addresses the knowledge and skill requirements of the operators to meet the task characteristics and the responses of the plant processes. The results to date and implications for going forward from this process are also presented. (authors)

  2. Bandages of boiled potato peels.

    Science.gov (United States)

    Patil, A R; Keswani, M H

    1985-08-01

    The use of potato peels as a dressing for burn wounds has been reported previously. A technique of preparing bandage rolls with boiled potato peels is now presented, which makes dressing of a burn wound more convenient. PMID:4041947

  3. High flux film and transition boiling

    Energy Technology Data Exchange (ETDEWEB)

    Witte, L.C.

    1990-01-01

    This report is a bench-scale experiment on transition boiling. The author gives a detailed description on experimental apparatus and conditions. The visual observed boiling phenomena; nucleate boiling and film boiling, and the effect of heat transfer are also elucidated. 10 refs., 11 figs., 1 tab.

  4. Enhanced heat transfer in confined pool boiling

    NARCIS (Netherlands)

    Rops, C.M.; Lindken, R.; Velthuis, J.F.M.; Westerweel, J.

    2009-01-01

    We report the results of an experimental investigation of the heat transfer during nucleate boiling on a spatially confined boiling surface. The heat flux as a function of the boiling surface temperature was measured in pool boiling pots with diameters ranging from 15 mm down to 4.5 mm. It was found

  5. Surface boiling of superheated liquid

    Energy Technology Data Exchange (ETDEWEB)

    Reinke, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-01-01

    A basic vaporization mechanism that possibly affects the qualitative and quantitative prediction of the consequences of accidental releases of hazardous superheated liquids was experimentally and analytically investigated. The studies are of relevance for the instantaneous failure of a containment vessel filled with liquefied gas. Even though catastrophical vessel failure is a rare event, it is considered to be a major technological hazard. Modeling the initial phase of depressurisation and vaporization of the contents is an essential step for the subsequent analysis of the spread and dispersion of the materials liberated. There is only limited understanding of this inertial expansion stage of the superheated liquid, before gravity and atmospheric turbulence begin to dominate the expansion. This work aims at a better understanding of the vaporization process and to supply more precise source-term data. It is also intended to provide knowledge for the prediction of the behavior of large-scale releases by the investigation of boiling on a small scale. Release experiments with butane, propane, R-134a and water were conducted. The vaporization of liquids that became superheated by sudden depressurisation was studied in nucleation-site-free glass receptacles. Several novel techniques for preventing undesired nucleation and for opening the test-section were developed. Releases from pipes and from a cylindrical geometry allowed both linear one-dimensional, and radial-front two-dimensional propagation to be investigated. Releases were made to atmospheric pressure over a range of superheats. It was found that, above a certain superheat temperature, the free surface of the metastable liquid rapidly broke up and ejected a high-velocity vapor/liquid stream. The zone of intense vaporization and liquid fragmentation proceeded as a front that advanced into the test fluids. No nucleation of bubbles in the bulk of the superheated liquid was observed. (author) figs., tabs., refs.

  6. Surface boiling of superheated liquid

    International Nuclear Information System (INIS)

    A basic vaporization mechanism that possibly affects the qualitative and quantitative prediction of the consequences of accidental releases of hazardous superheated liquids was experimentally and analytically investigated. The studies are of relevance for the instantaneous failure of a containment vessel filled with liquefied gas. Even though catastrophical vessel failure is a rare event, it is considered to be a major technological hazard. Modeling the initial phase of depressurisation and vaporization of the contents is an essential step for the subsequent analysis of the spread and dispersion of the materials liberated. There is only limited understanding of this inertial expansion stage of the superheated liquid, before gravity and atmospheric turbulence begin to dominate the expansion. This work aims at a better understanding of the vaporization process and to supply more precise source-term data. It is also intended to provide knowledge for the prediction of the behavior of large-scale releases by the investigation of boiling on a small scale. Release experiments with butane, propane, R-134a and water were conducted. The vaporization of liquids that became superheated by sudden depressurisation was studied in nucleation-site-free glass receptacles. Several novel techniques for preventing undesired nucleation and for opening the test-section were developed. Releases from pipes and from a cylindrical geometry allowed both linear one-dimensional, and radial-front two-dimensional propagation to be investigated. Releases were made to atmospheric pressure over a range of superheats. It was found that, above a certain superheat temperature, the free surface of the metastable liquid rapidly broke up and ejected a high-velocity vapor/liquid stream. The zone of intense vaporization and liquid fragmentation proceeded as a front that advanced into the test fluids. No nucleation of bubbles in the bulk of the superheated liquid was observed. (author) figs., tabs., refs

  7. LMFBR safety and sodium boiling

    Energy Technology Data Exchange (ETDEWEB)

    Hinkle, W.D.; Tschamper, P.M.; Fontana, M.H.; Henry, R.E.; Padilla, A. Jr.

    1978-01-01

    Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.

  8. Development boiling to sprinkled tube bundle

    Science.gov (United States)

    Kracík, Petr; Pospíšil, Jiří

    2016-03-01

    This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes' interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  9. The nucleate pool boiling dilemma

    International Nuclear Information System (INIS)

    It is shown that the scatter of experimental data is due to the history and machining finish of the heated surface. All experimental pool boiling data published to date, which does not specify precisely the characteristics of the heated surface cannot be expected to provide reliable design information. (U.K.)

  10. Numerical Modeling and Investigation of Boiling Phenomena

    OpenAIRE

    Kunkelmann, Christian

    2011-01-01

    The subject of the present thesis is the numerical modeling and investigation of boiling phenomena. The heat transfer during boiling is highly efficient and therefore used for many applications in power generation, process engineering and cooling of high performance electronics. The precise knowledge of particular boiling processes, their relevant parameters and limitations is of utmost importance for an optimized application. Therefore, the fundamentals of boiling heat transfer have been...

  11. Boiling heat transfer and droplet spreading of nanofluids.

    Science.gov (United States)

    Murshed, S M Sohel; de Castro, C A Nieto

    2013-11-01

    Nanofluids- a new class of heat transfer fluids have recently been a very attractive area of research due to their fascinating thermophysical properties and numerous potential benefits and applications in many important fields. However, there are many controversies and inconsistencies in reported arguments and experimental results on various thermal characteristics such as effective thermal conductivity, convective heat transfer coefficient and boiling heat transfer rate of nanofluids. As of today, researchers have mostly focused on anomalous thermal conductivity of nanofluids. Although investigations on boiling and droplet spreading are very important for practical application of nanofluids as advanced coolants, considerably fewer efforts have been made on these thermal features of nanofluids. In this paper, recent research and development in boiling heat transfer and droplet spreading of nanofluids are reviewed together with summarizing most related patents on nanofluids published in literature. Review reveals that despite some inconsistent results nanofluids exhibit significantly higher boiling heat transfer performance compared to their base fluids and show great promises to be used as advanced heat transfer fluids in numerous applications. However, there is a clear lack of in-depth understanding of heat transport mechanisms during phase change of nanofluids. It is also found that the nanofluids related patents are limited and among them most of the patents are based on thermal conductivity enhancement and synthesising processes of specific type of nanofluids.

  12. Advanced nuclear plants meet the economic challenge

    International Nuclear Information System (INIS)

    Nuclear power plants operated in the baseload regime are economically competitive even when compared with plants burning fossil fuels. As they do not produce emissions when operated, they do not pollute the environment. This is clearly reflected also in the internalized costs. After 2000, many new power plants are expected to be constructed in the USA and worldwide. An important role in this phase will be played by advanced light water reactors of the ABWR and SBWR types representing the future state of the art in technology and safety as well as in cost and plant operations management. (orig.)

  13. Burnout in boiling heat transfer. part I: pool boiling systems

    International Nuclear Information System (INIS)

    Recent experimental and analytical developments in pool-boiling burnout are reviewed, and results are summarized that clarify the dependence of critical heat flux on heater geometry and fluid properties. New analytical interpretations of burnout are discussed, and the effects of surface condition, aging, acceleration, and transient heating (or cooling) are described. The relation of sound to burnout and new techniques for stabilizing electric heaters at burnout are also considered

  14. Thermosyphon boiling in vertical channels

    Science.gov (United States)

    Bar-Cohen, A.; Schweitzer, H.

    The thermal characteristics of ebullient cooling systems for VHSIC and VLSI microelectronic component thermal control are studied by experimentally and analytically investigating boiling heat transfer from a pair of flat, closely spaced, isoflux plates immersed in saturated water. A theoretical model for liquid flow rate through the channel is developed and used as a basis for correlating the rate of heat transfer from the channel walls. Experimental results for wall temperature as a function of axial location, heat flux, and plate spacing are presented. The finding that the wall superheat at constant imposed heat flux decreases as the channel is narrowed is explained with the aid of a boiling thermosiphon analysis which yields the mass flux through the channel.

  15. Instability in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

    This Brief addresses the phenomena of instability in flow boiling in microchannels occurring in high heat flux electronic cooling. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Critical Heat Flux in Flow Boiling in Microchannels,” and "Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,"by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  16. Duality of boiling systems and uncertainty phenomena

    Institute of Scientific and Technical Information of China (English)

    柴立合; 彭晓峰; 王补宣

    2000-01-01

    Interactions among dry patches at high heat flux are theoretically analyzed. The high heat flux boiling experiments on metal plate wall with different materials and thickness are correspondingly conducted. The duality of boiling system, i.e. hydrodynamic performance and self-organized performance is identified. A unified explanation of hydrodynamic models and dry patches models is given. The scatter and uncertainty in boiling data can be mainly attributed to the intrinsic duality, but not the sole surface effects. The present experimental results explain why the deviation point at high flux boiling is seen only on occasion and why the self-organization of dry patches is often ignored in available literature.

  17. Boiling of the Interface between Two Immiscible Liquids below the Bulk Boiling Temperatures of Both Components

    OpenAIRE

    Pimenova, Anastasiya V.; Goldobin, Denis S.

    2014-01-01

    We consider the problem of boiling of the direct contact of two immiscible liquids. An intense vapour formation at such a direct contact is possible below the bulk boiling points of both components, meaning an effective decrease of the boiling temperature of the system. Although the phenomenon is known in science and widely employed in technology, the direct contact boiling process was thoroughly studied (both experimentally and theoretically) only for the case where one of liquids is becomin...

  18. Boiling flow through diverging microchannel

    Indian Academy of Sciences (India)

    V S Duryodhan; S G Singh; Amit Agrawal

    2013-12-01

    An experimental study of flow boiling through diverging microchannel has been carried out in this work, with the aim of understanding boiling in nonuniform cross-section microchannel. Diverging microchannel of 4° of divergence angle and 146 m hydraulic diameter (calculated at mid-length) has been employed for the present study with deionised water as working fluid. Effect of mass flux (118–1182 kg/m2-s) and heat flux (1.6–19.2 W/cm2) on single and two-phase pressure drop and average heat transfer coefficient has been studied. Concurrently, flow visualization is carried out to document the various flow regimes and to correlate the pressure drop and average heat transfer coefficient to the underlying flow regime. Four flow regimes have been identified from the measurements: bubbly, slug, slug–annular and periodic dry-out/rewetting. Variation of pressure drop with heat flux shows one maxima which corresponds to transition from bubbly to slug flow. It is shown that significantly large heat transfer coefficient (up to 107 kW/m2-K) can be attained for such systems, for small pressure drop penalty and with good flow stability.

  19. Boiling turbulent Rayleigh-Bénard convection

    NARCIS (Netherlands)

    Lakkaraju, R.

    2013-01-01

    A fundamental understanding of liquid-vapor phase transitions, mainly boiling phenomenon, is essential due to its omnipresence in science and technology. In industries, many empirical correlations exist on the heat transport to get an optimized and efficient thermal design of the boiling equipment.

  20. Transient boiling crisis of cryogenic liquids

    NARCIS (Netherlands)

    Deev, [No Value; Kharitonov, VS; Kutsenko, KV; Lavrukhin, AA

    2004-01-01

    This paper introduces a new physical model of boiling crisis under rapid increase of power on the heated surface. The calculation of the time interval of the transition to film boiling in cryogenic liquids was carried out depending on heat flux and pressure. The obtained results are in good agreemen

  1. Boiling heat transfer with acoustic cavitation

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The effects of acoustic cavitation and nanometer granule on boiling heat transfer of horizontal circular copper tube are investigated experimentally and theoretically using acetone as the working fluid according to the boiling procedure. The results show that heat transfer can be enhanced or weakened by generation of the cavitation bubble or addition of the nanometer granules, respectively. The mechanisms of the effects are analyzed.

  2. Transition boiling heat transfer during reflooding transients

    International Nuclear Information System (INIS)

    Transition boiling heat transfer is characterized by a heat flux which declines as the heater wall temperature increases. Steady state transition boiling is also characterized by alternate periods of high and low heat transfer caused by intermittent wetting of the heated surface. In flow boiling, the reason for intermittent wetting depends on the volume fraction of vapor present. At high vapor volume fractions, annular flow exists during what is generally called the nucleate boiling region, and a thin liquid film is present on the surface. The remainder of the passage is filled with vapor carrying entrained droplets. Above the nucleate boiling region there is no liquid film, and heat is transferred to droplet-laden vapor. In the narrow transition boiling region between nucleate boiling and heat transfer to steam, the liquid film is present only part of the time. The intermittent wetting produces significant wall temperature oscillations. Recent phenomenologically based modeling of steady state transition boiling heat transfer at high vapor fractions has been successful in predicting the magnitude of both temperature oscillations and heat transfer rates. After a brief review of the steady state model, this note shows how the results of the steady state analysis for vertical surfaces may be used to obtain heat transfer rates during reflooding transients

  3. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  4. Development of a mechanistic model for forced convection subcooled boiling

    Science.gov (United States)

    Shaver, Dillon R.

    The focus of this work is on the formulation, implementation, and testing of a mechanistic model of subcooled boiling. Subcooled boiling is the process of vapor generation on a heated wall when the bulk liquid temperature is still below saturation. This is part of a larger effort by the US DoE's CASL project to apply advanced computational tools to the simulation of light water reactors. To support this effort, the formulation of the dispersed field model is described and a complete model of interfacial forces is formulated. The model has been implemented in the NPHASE-CMFD computer code with a K-epsilon model of turbulence. The interfacial force models are built on extensive work by other authors, and include novel formulations of the turbulent dispersion and lift forces. The complete model of interfacial forces is compared to experiments for adiabatic bubbly flows, including both steady-state and unsteady conditions. The same model is then applied to a transient gas/liquid flow in a complex geometry of fuel channels in a sodium fast reactor. Building on the foundation of the interfacial force model, a mechanistic model of forced-convection subcooled boiling is proposed. This model uses the heat flux partitioning concept and accounts for condensation of bubbles attached to the wall. This allows the model to capture the enhanced heat transfer associated with boiling before the point of net generation of vapor, a phenomenon consistent with existing experimental observations. The model is compared to four different experiments encompassing flows of light water, heavy water, and R12 at different pressures, in cylindrical channels, an internally heated annulus, and a rectangular channel. The experimental data includes axial and radial profiles of both liquid temperature and vapor volume fraction, and the agreement can be considered quite good. The complete model is then applied to simulations of subcooled boiling in nuclear reactor subchannels consistent with the

  5. Boiling Heat Transfer to Halogenated Hydrocarbon Refrigerants

    Science.gov (United States)

    Yoshida, Suguru; Fujita, Yasunobu

    The current state of knowledge on heat transfer to boiling refrigerants (halogenated hydrocarbons) in a pool and flowing inside a horizontal tube is reviewed with an emphasis on information relevant to the design of refrigerant evaporators, and some recommendations are made for future research. The review covers two-phase flow pattern, heat transfer characteristics, correlation of heat transfer coefficient, influence of oil, heat transfer augmentation, boiling from tube-bundle, influence of return bend, burnout heat flux, film boiling, dryout and post-dryout heat transfer.

  6. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Tao [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)]. E-mail: zhoutao@mail.tsinghua.edu.cn; Wang Zenghui [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Yang Ruichang [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)

    2005-10-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well.

  7. Construction and preoperational test of Kashiwasaki-Kariwa Nuclear Power Station Unit No. 6

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan)

    1996-12-31

    Unit 6 of the Kashiwazaki-Kariwa nuclear power station of Tokyo Electric Power Company, the world`s first advanced boiling water reactor (ABWR), is progressing ahead of the originally established schedule since the start of its construction in September 1991, and commercial operation is scheduled to start before the end of 1996.

  8. 21 CFR 872.6710 - Boiling water sterilizer.

    Science.gov (United States)

    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...

  9. Using Boiling for Treating Waste Activated Sludge

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In this work we investigated the feasibility of using short time, low superheat boiling to treat biological sludge. The treated sludge exhibited reduced filterability and enhanced settleability. The boiling treatment released a large amount of extra-cellular polymers (ECPs) from the solid phase and reduced the microbial density levels of the total coliform bacteria and the heterotrophic bacteria. A diluted sludge is preferable for its high degree of organic hydrolysis and sufficient reduction in microbial density levels.

  10. How To Boil the Perfect Egg

    Institute of Scientific and Technical Information of China (English)

    小雨

    2007-01-01

    A British inventor says he has cracked(破解)the age-old riddle(难题)of how to boil the perfect egg,get rid of(摆脱)the water. Simon Rhymes uses powerful light bulbs instead of boiling water to cook the egg. The gadget(小发明)does the job in six minutes,and then chons off(削)the top of

  11. High flux film and transition boiling

    Science.gov (United States)

    Witte, L. C.

    1993-02-01

    An investigation was conducted on the potential for altering the boiling curve through effects of high velocity and high subcooling. Experiments using water and Freon-113 flowing over cylindrical electrical heaters in crossflow were made to see how velocity and subcooling affect the boiling curve, especially the film and transition boiling regions. We sought subcooling levels down to near the freezing points of these two liquids to prove the concept that the critical heat flux and the minimum heat flux could be brought together, thereby averting the transition region altogether. Another emphasis was to gain insight into how the various boiling regions could be represented mathematically on various parts of the heating surface. Motivation for the research grew out of a realization that the effects of very high subcooling and velocity might be to avert the transition boiling altogether so that the unstable part of the boiling curve would not limit the application of high flux devices to temperatures less than the burnout temperatures. Summaries of results from the study are described. It shows that the potential for averting the transition region is good and points the way to further research that is needed to demonstrate the potential.

  12. Lattice Boltzmann modeling of boiling heat transfer: The boiling curve and the effects of wettability

    CERN Document Server

    Li, Q; Francois, M M; He, Y L; Luo, K H

    2015-01-01

    A hybrid thermal lattice Boltzmann (LB) model is presented to simulate thermal multiphase flows with phase change based on an improved pseudopotential LB approach [Q. Li, K. H. Luo, and X. J. Li, Phys. Rev. E 87, 053301 (2013)]. The present model does not suffer from the spurious term caused by the forcing-term effect, which was encountered in some previous thermal LB models for liquid-vapor phase change. Using the model, the liquid-vapor boiling process is simulated. The boiling curve together with the three boiling stages (nucleate boiling, transition boiling, and film boiling) is numerically reproduced in the LB community for the first time. The numerical results show that the basic features and the fundamental characteristics of boiling heat transfer are well captured, such as the severe fluctuation of transient heat flux in the transition boiling and the feature that the maximum heat transfer coefficient lies at a lower wall superheat than that of the maximum heat flux. Furthermore, the effects of the he...

  13. Structural Changes of Malt Proteins During Boiling

    Directory of Open Access Journals (Sweden)

    2009-03-01

    Full Text Available Changes in the physicochemical properties and structure of proteins derived from two malt varieties (Baudin and Guangmai during wort boiling were investigated by differential scanning calorimetry, SDS-PAGE, two-dimensional electrophoresis, gel filtration chromatography and circular dichroism spectroscopy. The results showed that both protein content and amino acid composition changed only slightly during boiling, and that boiling might cause a gradual unfolding of protein structures, as indicated by the decrease in surface hydrophobicity and free sulfhydryl content and enthalpy value, as well as reduced α-helix contents and markedly increased random coil contents. It was also found that major component of both worts was a boiling-resistant protein with a molecular mass of 40 kDa, and that according to the two-dimensional electrophoresis and SE-HPLC analyses, a small amount of soluble aggregates might be formed via hydrophobic interactions. It was thus concluded that changes of protein structure caused by boiling that might influence beer quality are largely independent of malt variety.

  14. How does surface wettability influence nucleate boiling?

    Science.gov (United States)

    Phan, Hai Trieu; Caney, Nadia; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2009-05-01

    Although the boiling process has been a major subject of research for several decades, its physics still remain unclear and require further investigation. This study aims at highlighting the effects of surface wettability on pool boiling heat transfer. Nanocoating techniques were used to vary the water contact angle from 20° to 110° by modifying nanoscale surface topography and chemistry. The experimental results obtained disagree with the predictions of the classical models. A new approach of nucleation mechanism is established to clarify the nexus between the surface wettability and the nucleate boiling heat transfer. In this approach, we introduce the concept of macro- and micro-contact angles to explain the observed phenomenon. To cite this article: H.T. Phan et al., C. R. Mecanique 337 (2009).

  15. The entropy balance for boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco-Javier E-mail: fjk@posta.unizar.es

    2001-10-01

    Subcooled forced convection boiling of water is recognized as one of the best means of accommodating the very high heat fluxes that plasma facing components of fusion reactors have to withstand. The boiling curve, giving the wall temperature in function of the applied flux and flow conditions, is essential for the design of such cooling configurations. In this paper, a new entropy balance for subcooled boiling flow, which allows the wall temperature to be obtained, is presented and successfully compared with experimental data from the Joint US-EURATOM R and D Program. The derivation of this entropy balance is based on a new strict application of the Reynolds theorem to multiphase flows recently proposed by the author.

  16. Thermodynamics of Flow Boiling Heat Transfer

    Science.gov (United States)

    Collado, F. J.

    2003-05-01

    Convective boiling in sub-cooled water flowing through a heated channel is essential in many engineering applications where high heat flux needs to be accommodated. It has been customary to represent the heat transfer by the boiling curve, which shows the heat flux versus the wall-minus-saturation temperature difference. However it is a rather complicated problem, and recent revisions of two-phase flow and heat transfer note that calculated values of boiling heat transfer coefficients present many uncertainties. Quite recently, the author has shown that the average thermal gap in the heated channel (the wall temperature minus the average temperature of the coolant) was tightly connected with the thermodynamic efficiency of a theoretical reversible engine placed in this thermal gap. In this work, whereas this correlation is checked again with data taken by General Electric (task III) for water at high pressure, a possible connection between this wall efficiency and the reversible-work theorem is explored.

  17. The boiling point of stratospheric aerosols.

    Science.gov (United States)

    Rosen, J. M.

    1971-01-01

    A photoelectric particle counter was used for the measurement of aerosol boiling points. The operational principle involves raising the temperature of the aerosol by vigorously heating a portion of the intake tube. At or above the boiling point, the particles disintegrate rather quickly, and a noticeable effect on the size distribution and concentration is observed. Stratospheric aerosols appear to have the same volatility as a solution of 75% sulfuric acid. Chemical analysis of the aerosols indicates that there are other substances present, but that the sulfate radical is apparently the major constituent.

  18. Water boiling kinetic in rapid decompression

    International Nuclear Information System (INIS)

    This study entering in the frame of a CEA, EDF and Framatome collaboration, has for objective to modelize two-phase flows in case of PWR Loca. The objective is to find, by taking in account the all imbalances, a formulation for the mass transfer at the interface water-vapor by the study of water boiling phenomenon in case of fast decompression such as a primary circuit break. In this accident, the estimation of boiling speeds in an essential parameter for determining the break discharge which conditions the safety systems design

  19. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  20. Cryogenic Propellant Boil-Off Reduction System

    Science.gov (United States)

    Plachta, D. W.; Christie, R. J.; Carlberg, E.; Feller, J. R.

    2008-03-01

    Lunar missions under consideration would benefit from incorporation of high specific impulse propellants such as LH2 and LO2, even with their accompanying boil-off losses necessary to maintain a steady tank pressure. This paper addresses a cryogenic propellant boil-off reduction system to minimize or eliminate boil-off. Concepts to do so were considered under the In-Space Cryogenic Propellant Depot Project. Specific to that was an investigation of cryocooler integration concepts for relatively large depot sized propellant tanks. One concept proved promising—it served to efficiently move heat to the cryocooler even over long distances via a compressed helium loop. The analyses and designs for this were incorporated into NASA Glenn Research Center's Cryogenic Analysis Tool. That design approach is explained and shown herein. Analysis shows that, when compared to passive only cryogenic storage, the boil-off reduction system begins to reduce system mass if durations are as low as 40 days for LH2, and 14 days for LO2. In addition, a method of cooling LH2 tanks is presented that precludes development issues associated with LH2 temperature cryocoolers.

  1. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  2. Classic and Hard-Boiled Detective Fiction.

    Science.gov (United States)

    Reilly, John M.

    Through an analysis of several stories, this paper defines the similarities and differences between classic and hard-boiled detective fiction. The characters and plots of three stories are discussed: "The Red House" by A. A. Milne; "I, The Jury" by Mickey Spillane; and "League of Frightened Men" by Rex Stout. The classic detective story is defined…

  3. Heat transfer coefficient for boiling carbon dioxide

    DEFF Research Database (Denmark)

    Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

    1998-01-01

    Heat transfer coefficient and pressure drop for boiling carbon dioxide (R744) flowing in a horizontal pipe has been measured. The calculated heat transfer coeeficient has been compared with the Chart correlation of Shah. The Chart Correlation predits too low heat transfer coefficient but the ratio...

  4. Heat transfer coeffcient for boiling carbon dioxide

    DEFF Research Database (Denmark)

    Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

    1997-01-01

    Heat transfer coefficient and pressure drop for boiling carbon dioxide (R744) flowing in a horizontal pipe has been measured. The pipe is heated by condensing R22 outside the pipe. The heat input is supplied by an electrical heater wich evaporates the R22. With the heat flux assumed constant over...

  5. The law of stable equilibrium and the entropy-based boiling curve for flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Collado, F.J. [Universidad de Zaragoza (Spain). Dpto. Ingenieria Mecanica Motores Termicos

    2005-05-01

    Convective flow boiling in sub-cooled fluids is recognized as one of the few means of accommodating very high heat fluxes. There are many available correlations for predicting the inner wall temperature of the heated duct in the several regimes of the empirical Nukiyama boiling curve, although unfortunately there is no physical fundamentals of such curve. Recently, the author has shown that the classical entropy balance could contain key information about boiling heat transfer. So, it was found that the average thermal gap in the heated channel (the inner wall temperature minus the average temperature of the coolant fluid) was strongly correlated with the efficiency of a theoretical reversible engine placed in this thermal gap. From this new correlation, a new boiling curve plotting the wall temperature versus the average fluid temperature was derived and successfully checked against low- and high-pressure water data. This curve suggested a new and simple definition of the critical heat flux (CHF) namely, the value of the coolant average temperature at the maximum. In this work, after briefly reviewing the entropy balance of a non-equilibrium boiling flow and its relationship with the thermodynamic average temperature and the law of stable equilibrium (LSE), the possibilities of the new approach for the design of flow boiling cooling systems are highlighted. Finally, the strong correlation found between the reversible engine efficiency and the thermal driving force is verified again, now with high-pressure refrigerant 22 (R-22) data. (author)

  6. Analytical simulation of boiling water reactor pressure suppression pool swell

    International Nuclear Information System (INIS)

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement

  7. Analytical simulation of boiling water reactor pressure suppression pool swell

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

  8. Flow boiling in microgap channels experiment, visualization and analysis

    CERN Document Server

    Alam, Tamanna; Jin, Li-Wen

    2013-01-01

    Flow Boiling in Microgap Channels: Experiment, Visualization and Analysis presents an up-to-date summary of the details of the confined to unconfined flow boiling transition criteria, flow boiling heat transfer and pressure drop characteristics, instability characteristics, two phase flow pattern and flow regime map and the parametric study of microgap dimension. Advantages of flow boiling in microgaps over microchannels are also highlighted. The objective of this Brief is to obtain a better fundamental understanding of the flow boiling processes, compare the performance between microgap and c

  9. Dynamic Bubble Behaviour during Microscale Subcooled Boiling

    Institute of Scientific and Technical Information of China (English)

    WANG Hao; PENG Xiao-Feng; David M.Christopher

    2005-01-01

    @@ Bubble cycles, including initiation, growth and departure, are the physical basis of nucleate boiling. The presentinvestigation, however, reveals unusual bubble motions during subcooled nucleate boiling on microwires 25 orl00μm in diameter. Two types of bubble motions, bubble sweeping and bubble return, are observed in theexperiments. Bubble sweeping describes a bubble moving back and forth along the wire, which is motion parallelto the wire. Bubble return is the bubble moving back to the wire after it has detached or leaping above thewire. Theoretical analyses and numerical simulations are conducted to investigate the driving mechanisms forboth bubble sweeping and return. Marangoni flow from warm to cool regions along the bubble interface is foundto produce the shear stresses needed to drive these unusual bubble movements.

  10. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    张利斌; 李修伦

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39 mm ID and 2.0 m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum.The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  11. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39mm ID and 2.0m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum. The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  12. Enzyme engineering reaches the boiling point

    OpenAIRE

    Arnold, Frances H.

    1998-01-01

    The boiled enzyme was toppled as a standard enzymology control when researchers in the 1970s started uncovering enzymes that loved the heat (1). Identification of a variety of intrinsically hyperstable enzymes from hyperthermophilic organisms, with optimal growth temperatures of 100°C and above, has piqued academic curiosity (e.g., how do these proteins withstand such ‘‘extreme’’ conditions?) and generated considerable interest for their possible applications in biotechnology (2, 3). The real...

  13. European simplified boiling water reactor (ESBWR) plant

    International Nuclear Information System (INIS)

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility

  14. Self-propelled film-boiling liquids

    CERN Document Server

    Linke, H; Melling, L D; Taormina, M J; Francis, M J; Dow-Hygelund, C C; Narayanan, V K; Taylor, R P; Stout, A

    2005-01-01

    We report that liquids perform self-propelled motion when they are placed in contact with hot surfaces with asymmetric (ratchet-like) topology. The pumping effect is observed when the liquid is in the film-boiling regime, for many liquids and over a wide temperature range. We propose that liquid motion is driven by a viscous force exerted by vapor flow between the solid and the liquid.

  15. A study of flow boiling phenomena using real time neutron radiography

    Science.gov (United States)

    Novog, David Raymond

    The operation and safety of both fossil-fuel and nuclear power stations depend on adequate cooling of the thermal source involved. This is usually accomplished using liquid coolants that are forced through the high temperature regions by a pumping system; this fluid then transports the thermal energy to another section of the power station. However, fluids that undergo boiling during this process create vapor that can be detrimental, and influence safe operation of other system components. The behavior of this vapor, or void, as it is generated and transported through the system is critical in predicting the operational and safety performance. This study uses two advanced penetrating radiation techniques, Real Time Neutron Radiography (RTNR), and High Speed X-Ray Tomography (HS-XCT), to examine void generation and transport behavior in a flow boiling system. The geometries studied were tube side flow boiling in a cylindrical configuration, and a similar flow channel with an internal twisted tape swirl flow generator. The heat transfer performance and pressure drop characteristics were monitored in addition to void distribution measurements, so that the impact of void distribution could be determined. The RTNR and heat transfer pipe flow studies were conducted using boiling Refrigerant 134a at pressures from 500 to 700 kPa, inlet subcooling from 3 to 12°C and mass fluxes from 55 to 170kg/m 2-s with heat fluxes up to 40 kW/m2. RTNR and HS-XCT were used to measure the distribution and size of the vapor phases in the channel for cylindrical tube-side flow boiling and swirl-flow boiling geometries. The results clearly show that the averaged void is similar for both geometries, but that there is a significant difference in the void distribution, velocity and transport behavior from one configuration to the next. Specifically, the void distribution during flow boiling in a cylindrical-tube test section showed that the void fraction was largest near the tube center and

  16. CFD for Subcooled Flow Boiling: Parametric Variations

    Directory of Open Access Journals (Sweden)

    Roland Rzehak

    2013-01-01

    Full Text Available We investigate the present capabilities of CFD for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Very similar modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant nondimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12 as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, liquid temperature, and bubble size. Robust predictive capabilities of the modeling require that it is validated for a wide range of parameters. It is known that a careful calibration of correlations used in the wall boiling model is necessary to obtain agreement with the measured data. We here consider tests under a variety of conditions concerning liquid subcooling, flow rate, and heat flux. It is investigated to which extent a set of calibrated model parameters suffices to cover at least a certain parameter range.

  17. Pressure drop in saturated flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco J. [Universidad de Zaragoza, Zaragoza (Spain)

    2003-07-01

    A new mass balance for flow boiling have been recently suggested by the author following a quite simple idea: if the phases have different velocities, they can not cover the same distance -the control volume length for a 1-d system- in the same time. Thus, the time scales of the phases have to be different, and we should scale the time dependent magnitudes of one phase to the other one before combining them. Furthermore, it is reasonable to think that conservation equations should have to include in some manner this evident physical fact. In complete coherence with the former mass balance, a new energy balance, which does include the slip ratio has been also stated. This work, whilst reviews these new fundamentals for saturated flow boiling, stresses those aspects related with the prediction of the pressure drop in saturated flow boiling. The new correlations found for the data carefully measured by Thom during the Cambridge project would confirm the new two-phase flowapproach.

  18. Boiling of an Emulsion in a Yield Stress Fluid

    OpenAIRE

    Guéna, Geoffroy; Wang, Ji; D'Espinose, Jean-Baptiste; Lequeux, François; Talini, Laurence

    2010-01-01

    International audience We report the boiling behaviour of pentane emulsified in a yield stress fluid, a colloidal clay (Laponite) suspension. We have observed that a superheated state is easily reached: the emulsion, heated more than 50°C above the alkane boiling point, does not boil. Superheating is made possible by the suppression of heterogeneous nucleation in pentane, resulting from the emulsification process, a phenomenon evidenced decades ago in studies of the superheating of two pha...

  19. Water Boiling inside Carbon Nanotubes: Towards Efficient Drug Release

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Oleg V.

    2012-01-01

    We show using molecular dynamics simulation that spatial confinement of water inside carbon nanotubes (CNT) 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 trans...

  20. Prospective Chemistry Teachers’ Understanding of Boiling: A Phenomenological Study

    OpenAIRE

    CANPOLAT, Nurtaç; PINARBAŞI, Tacettin

    2012-01-01

    This study investigates chemistry prospective teachers’ views regarding boiling phenomenon, and provides a concept analysis on the nature of boiling together with suggestions on how to teach boiling phenomenon in the light of literature and findings of this study. The sample of this study consists of 18 senior prospective chemistry teachers who attend chemistry teacher training program. Data were collected by discussions with the participants. The discussions were specifically focused on pros...

  1. Explosive Boiling of Superheated Cryogenic Liquids

    CERN Document Server

    Baidakov, V G

    2007-01-01

    The monograph is devoted to the description of the kinetics of spontaneous boiling of superheated liquefied gases and their solutions. Experimental results are given on the temperature of accessible superheating, the limits of tensile strength of liquids due to processes of cavitation and the rates of nucleation of classical and quantum liquids. The kinetics of evolution of the gas phase is studied in detail for solutions of cryogenic liquids and gas-saturated fluids. The properties of the critical clusters (bubbles of critical sizes) of the newly evolving gas phase are analyzed for initial st

  2. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    This master thesis describes a mathematical model of a boiling water reactor and address the dynamic behaviour of the neutron kinetics, boilding dynamics and pressur stability. The simulation have been done using the SIMNON-program. The meaning were that the result from this work possibly would be adjust to supervision methods suitable for application in computer systems. This master thesis in automatic control has been done at the Department of Automatic Control, Lund Institute of Technology. The initiative to the work came from Sydkraft AB. (author)

  3. Interfacial wavy motion during film boiling from a downward-facing curved surface

    International Nuclear Information System (INIS)

    In the process of designing for the APR1400(Advanced Power Reactor 1400 MWe, the concept of in-vessel retention through external vessel cooling(IVR-EVC) was chosen as a severe accident management strategy. The cavity flooding was selected as the external vessel cooling method because of simpler installation relative to flooding within the thermal insulator. In fact, the IVR-EVC concept had not been considered during the initial design phase of the APR1400. Thus, several issues surfaced while applying the IVR concept at a later stage of design. One of these issues centered about delayed flooding of the reactor vessel because of the large volume between the cavity floor and the lower head. The cavity flooding may take as much as forty minutes depending upon the accidents scenario. It is thus not certain whether the flooding time will always be shorter than the time for relocation of the molten core material to the lower plenum of the reactor vessel. In addition, the initial temperature of the vessel, which should be in the vicinity of the saturation point corresponding to the primary system pressure, will far exceed temperature of the cavity flooding water during an accident. Hence, the initial hear removal mechanism for external vessel cooling will most likely be film rather than nucleate boiling. The results of this work indicate, however, that film boiling heat transfer coefficients presently available in the literature tend to underpredict the actual value for the reactor vessel lower head. In this study, In this study, film boiling heat transfer coefficients are obtained from the DELTA(Downward-boiling Experiment Laminar Transition Apparatus) quenching test utilizing the measured temperature histories. They are compared with the other experiment of the same edge angle. The film boiling heat transfer phenomena are visualized through a digital camera

  4. Pool boiling heat transfer performance of Newtonian nanofluids

    Energy Technology Data Exchange (ETDEWEB)

    Soltani, Saide; Etemad, Seyed Gholamreza [Isfahan University of Technology, Department of Chemical Engineering, Isfahan (Iran); Thibault, Jules [University of Ottawa, Department of Chemical and Biological Engineering, Ottawa, ON (Canada)

    2009-10-15

    Experimental measurements were carried out on the boiling heat transfer characteristics of {gamma}-Al{sub 2}O{sub 3}/water and SnO{sub 2}/water Newtonian nanofluids. Nanofluids are liquid suspensions containing nanoparticles with sizes smaller than 100 nm. In this research, suspensions with different concentrations of {gamma}-Al{sub 2}O{sub 3} and SnO{sub 2} nanoparticles in water were studied under nucleate pool boiling heat transfer conditions. Results show that nanofluids possess noticeably higher boiling heat transfer coefficients than the base fluid. The boiling heat transfer coefficients depend on the type and concentration of nanoparticles. (orig.)

  5. Investigation of Enhanced Boiling Heat Transfer from Porous Surfaces

    Institute of Scientific and Technical Information of China (English)

    LinZhiping; MaTongze; 等

    1994-01-01

    Experimental investigations of boiling heat transfer from porous surfaces at atmospheric pressure were performne.The porous surfaces are plain tubes coverd with metal screens.V-shaped groove tubes covered with screens,plain tubes sintered with screens.and V-shaped groove tubes sintered with screens,The experimental results show that sintering metal screens around spiral V-shaped groove tubes can greatly improve the boiling heat transfer,The boiling hystesis was observed in the experiment.This paper discusses the mechanism of the boiling heat transfer from those kinds of porous surfaces stated above.

  6. Zero Boil-Off Tank (ZBOT) Experiment

    Science.gov (United States)

    Mcquillen, John

    2016-01-01

    The Zero-Boil-Off Tank (ZBOT) experiment has been developed as a small scale ISS experiment aimed at delineating important fluid flow, heat and mass transport, and phase change phenomena that affect cryogenic storage tank pressurization and pressure control in microgravity. The experiments use a simulant transparent low boiling point fluid (PnP) in a sealed transparent Dewar to study and quantify: (a) fluid flow and thermal stratification during pressurization; (b) mixing, thermal destratification, depressurization, and jet-ullage penetration during pressure control by jet mixing. The experiment will provide valuable microgravity empirical two-phase data associated with the above-mentioned physical phenomena through highly accurate local wall and fluid temperature and pressure measurements, full-field phase-distribution and flow visualization. Moreover, the experiments are performed under tightly controlled and definable heat transfer boundary conditions to provide reliable high-fidelity data and precise input as required for validation verification of state-of-the-art two-phase CFD models developed as part of this research and by other groups in the international scientific and cryogenic fluid management communities.

  7. Mitigation performance indicator for boiling water reactors

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), and most currently use or plan to use noble metals technology. The EPRI Boiling Water Reactor Vessels and Internals Project (BWRVIP) developed a Mitigation Performance Indicator (MPI) in 2006 to accurately depict to management the status of mitigation equipment and as a standardized way to show the overall health of reactor vessel internals from a chemistry perspective. It is a 'Needed' requirement in the EPRI BWR Water Chemistry Guidelines that plants have an MPI, and use of the BWRVIP MPI is a 'Good Practice'. The MPI is aligned with inspection relief criteria for reactor piping and internal components for U.S. BWRs. This paper discusses the history of the MPI, from its first use for plants operating with moderate hydrogen water chemistry (HWC-M) or Noble Metal Chemical Application (NMCA) + HWC to its more recent use for plants operating with On-Line NobleChem™ (OLNC) + HWC. Key mitigation parameters are discussed along with the technical bases for the indicators associated with the parameters. (author)

  8. Zero boil-off system testing

    Science.gov (United States)

    Plachta, D. W.; Johnson, W. L.; Feller, J. R.

    2016-03-01

    Cryogenic propellants such as liquid hydrogen (LH2) and liquid oxygen (LO2) are a part of NASA's future space exploration plans due to their high specific impulse for rocket motors of upper stages. However, the low storage temperatures of LH2 and LO2 cause substantial boil-off losses for long duration missions. These losses can be eliminated by incorporating high performance cryocooler technology to intercept heat load to the propellant tanks and modulating the cryocooler temperature to control tank pressure. The technology being developed by NASA is the reverse turbo-Brayton cycle cryocooler and its integration to the propellant tank through a distributed cooling tubing network coupled to the tank wall. This configuration was recently tested at NASA Glenn Research Center in a vacuum chamber and cryoshroud that simulated the essential thermal aspects of low Earth orbit, its vacuum and temperature. This test series established that the active cooling system integrated with the propellant tank eliminated boil-off and robustly controlled tank pressure.

  9. Unsteady heat transfer during subcooled film boiling

    Science.gov (United States)

    Yagov, V. V.; Zabirov, A. R.; Lexin, M. A.

    2015-11-01

    Cooling of high-temperature bodies in subcooled liquid is of importance for quenching technologies and also for understanding the processes initiating vapor explosion. An analysis of the available experimental information shows that the mechanisms governing heat transfer in these processes are interpreted ambiguously; a more clear-cut definition of the Leidenfrost temperature notion is required. The results of experimental observations (Hewitt, Kenning, and previous investigations performed by the authors of this article) allow us to draw a conclusion that there exists a special mode of intense heat transfer during film boil- ing of highly subcooled liquid. For revealing regularities and mechanisms governing intense transfer of energy in this process, specialists of Moscow Power Engineering Institute's (MPEI) Department of Engineering Thermal Physics conduct systematic works aimed at investigating the cooling of high-temperature balls made of different metals in water with a temperature ranging from 20 to 100°C. It has been determined that the field of temperatures that takes place in balls with a diameter of more than 30 mm in intense cooling modes loses its spherical symmetry. An approximate procedure for solving the inverse thermal conductivity problem for calculating the heat flux density on the ball surface is developed. During film boiling, in which the ball surface temperature is well above the critical level for water, and in which liquid cannot come in direct contact with the wall, the calculated heat fluxes reach 3-7 MW/m2.

  10. A survey on the development of advanced instrumentation and control system in NPP

    International Nuclear Information System (INIS)

    Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: 1) N4 of EDF in France, 2) AP 600 of Westinghouse in USA, 3) NUPLEX-80+ of ABB-CE in USA, 4) CANDU in Canada, 5) Ohi 3 and 4, APWR and ABWR in Japan, 6) Belt-D in Germany, 7) Sizewell B in Britain, 8) Halden Reactor Projector in Norway, 9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author)

  11. A survey on the development of advanced instrumentation and control system in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: (1) N4 of EDF in France, (2) AP 600 of Westinghouse in USA, (3) NUPLEX-80+ of ABB-CE in USA, (4) CANDU in Canada, (5) Ohi 3 and 4, APWR and ABWR in Japan, (6) Belt-D in Germany, (7) Sizewell B in Britain, (8) Halden Reactor Projector in Norway, (9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author).

  12. Low-Flow Film Boiling Heat Transfer on Vertical Surfaces

    DEFF Research Database (Denmark)

    Munthe Andersen, J. G.; Dix, G. E.; Leonard, J. E.;

    1976-01-01

    The phenomenon of film boiling heat transfer for high wall temperatures has been investigated. Based on the assumption of laminar flow for the film, the continuity, momentum, and energy equations for the vapor film are solved and a Bromley-type analytical expression for the heat transfer...... length, an average film boiling heat transfer coefficient is obtained....

  13. Direct Numerical Simulation and Visualization of Subcooled Pool Boiling

    Directory of Open Access Journals (Sweden)

    Tomoaki Kunugi

    2014-01-01

    Full Text Available A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors. On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated.

  14. Technique for technological calculation of critical flow of boiling water

    International Nuclear Information System (INIS)

    Average values of friction factor and mach number for a critical flow of boiling water are determined on the basis of computerized processing of experimental data. Empirical formula, relating these values, which can be used for technological calculations of critical conditions of boiling water flow through transport pipelines, is derived

  15. Bubble transport in subcooled flow boiling

    Science.gov (United States)

    Owoeye, Eyitayo James

    Understanding the behavior of bubbles in subcooled flow boiling is important for optimum design and safety in several industrial applications. Bubble dynamics involve a complex combination of multiphase flow, heat transfer, and turbulence. When a vapor bubble is nucleated on a vertical heated wall, it typically slides and grows along the wall until it detaches into the bulk liquid. The bubble transfers heat from the wall into the subcooled liquid during this process. Effective control of this transport phenomenon is important for nuclear reactor cooling and requires the study of interfacial heat and mass transfer in a turbulent flow. Three approaches are commonly used in computational analysis of two-phase flow: Eulerian-Lagrangian, Eulerian-Eulerian, and interface tracking methods. The Eulerian- Lagrangian model assumes a spherical non-deformable bubble in a homogeneous domain. The Eulerian-Eulerian model solves separate conservation equations for each phase using averaging and closure laws. The interface tracking method solves a single set of conservation equations with the interfacial properties computed from the properties of both phases. It is less computationally expensive and does not require empirical relations at the fluid interface. Among the most established interface tracking techniques is the volume-of-fluid (VOF) method. VOF is accurate, conserves mass, captures topology changes, and permits sharp interfaces. This work involves the behavior of vapor bubbles in upward subcooled flow boiling. Both laminar and turbulent flow conditions are considered with corresponding pipe Reynolds number of 0 -- 410,000 using a large eddy simulation (LES) turbulence model and VOF interface tracking method. The study was performed at operating conditions that cover those of boiling water reactors (BWR) and pressurized water reactors (PWR). The analysis focused on the life cycle of vapor bubble after departing from its nucleation site, i.e. growth, slide, lift-off, rise

  16. CFD modelling of subcooled flow boiling for nuclear engineering applications

    International Nuclear Information System (INIS)

    In this work a general-purpose CFD code CFX-5 was used for simulations of subcooled flow boiling. The subcooled boiling model, available in a custom version of CFX-5, uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data [1] of high-pressure flow boiling in a vertical pipe covering a wide range of conditions (relevant to the pressurized water reactor). In general, a good agreement with the experimental data has been achieved. To adequately predict the lateral distribution of two-phase flow parameters, the modelling of two-phase flow turbulence and non-drag forces under wall boiling conditions have been also investigated in the paper. (author)

  17. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    OpenAIRE

    Nikolayev, Vadim; Beysens, D.; Garrabos, Yves; Lecoutre, Carole; Chatain, D.

    2006-01-01

    International audience Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil ...

  18. The Physics of Boiling at Burnout

    Science.gov (United States)

    Theofanous, T. G.; Tu, J. P.; Dinh, T. N.; Salmassi, T.; Dinh, A. T.; Gasljevic, K.

    2000-01-01

    The basic elements of a new experimental approach for the investigation of burnout in pool boiling are presented. The approach consists of the combined use of ultrathin (nano-scale) heaters and high speed infrared imaging of the heater temperature pattern as a whole, in conjunction with highly detailed control and characterization of heater morphology at the nano and micron scales. It is shown that the burnout phenomenon can be resolved in both space and time. Ultrathin heaters capable of dissipating power levels, at steady-state, of over 1 MW/square m are demonstrated. A separation of scales is identified and it is used to transfer the focus of attention from the complexity of the two-phase mixing layer in the vicinity of the heater to a micron-scaled microlayer and nucleation and associated film-disruption processes within it.

  19. Self-propelled film-boiling liquids

    Science.gov (United States)

    Linke, Heiner; Taormina, Michael; Aleman, Benjamin; Melling, Laura; Dow-Hygelund, Corey; Taylor, Richard; Francis, Matthew

    2006-03-01

    We report that liquids perform self-propelled motion when they are placed in contact with hot surfaces with asymmetric (ratchet-like) topology. Millimeter-sized droplets or slugs accelerate at rates up to 0.1 g and reach terminal velocities of several cm/s, sustained over distances up to a meter. The pumping effect is observed when the liquid is in the film-boiling regime, for many liquids and over a wide temperature range. We propose that liquid motion is driven by a viscous force exerted by vapor flow between the solid and the liquid. This heat-driven pumping mechanism may be of interest in cooling applications, eliminating the need for an additional power source.

  20. Hybrid modelling of a sugar boiling process

    CERN Document Server

    Lauret, Alfred Jean Philippe; Gatina, Jean Claude

    2012-01-01

    The first and maybe the most important step in designing a model-based predictive controller is to develop a model that is as accurate as possible and that is valid under a wide range of operating conditions. The sugar boiling process is a strongly nonlinear and nonstationary process. The main process nonlinearities are represented by the crystal growth rate. This paper addresses the development of the crystal growth rate model according to two approaches. The first approach is classical and consists of determining the parameters of the empirical expressions of the growth rate through the use of a nonlinear programming optimization technique. The second is a novel modeling strategy that combines an artificial neural network (ANN) as an approximator of the growth rate with prior knowledge represented by the mass balance of sucrose crystals. The first results show that the first type of model performs local fitting while the second offers a greater flexibility. The two models were developed with industrial data...

  1. High level disinfection of a home care device; to boil or not to boil?

    Science.gov (United States)

    Winthrop, K L; Homestead, N

    2012-03-01

    We developed a percutaneous electrical transducer for home therapy of chronic pain, a device that requires high level disinfection between uses. The utility of boiling water to provide high level disinfection was evaluated by inoculating transducer pads with potential skin pathogens (Staphylococcus aureus, Mycobacterium terrae, Pseudomonas aeruginosa, Candida albicans) and subjecting them to full immersion in water boiling at 4200 feet elevation (95 °C). Log10 reductions in colony-forming units (cfu) at 10 min were 7.1, >6.3 and >5.5 for S. aureus, P. aeruginosa and C. albicans, respectively, but only 4.6 for M. terrae. At 15 min the reductions had increased to 7.5, >6.8, >6.6 and >7.5 cfu, respectively.

  2. A contribution to incipient boiling in the case of subcooled boiling with forced convection

    International Nuclear Information System (INIS)

    The literature gives contradictory statements about incipient subcooled boiling. To clear up these contradictions it seems important to study the effect of different thermo- and hydrodynamic parameters, like heating surface load, system pressure, local supercooling, and flowrate. Further influencing quantities investigated here are the concentration dissolved gases and the surface condition of the heat surface. To carry out the experimental investigations a measuring method which has already been used by Mayinger applied. With this method, incipient boiling can be determined as the first measurable heat transfer improvement in comparison with single-phase forced convection. Besides, photographs sould make it possible to give statements on the quantity and size of the bubbles on the heating surface. (orig./GL)

  3. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  4. Micro-bubble emission boiling with the cavitation bubble blow pit

    Science.gov (United States)

    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

  5. Micro-bubble emission boiling with the cavitation bubble blow pit.

    Science.gov (United States)

    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

  6. Micro-bubble emission boiling with the cavitation bubble blow pit

    Science.gov (United States)

    Inada, Shigeaki; Shinagawa, Kazuaki; Illias, Suhaimi Bin; Sumiya, Hiroyuki; Jalaludin, Helmisyah A.

    2016-09-01

    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.

  7. Effects of structural parameters on flow boiling performance of reentrant porous microchannels

    Science.gov (United States)

    Deng, Daxiang; Tang, Yong; Shao, Haoran; Zeng, Jian; Zhou, Wei; Liang, Dejie

    2014-06-01

    Flow boiling within advanced microchannel heat sinks provides an efficient and attractive method for the cooling of microelectronics chips. In this study, a series of porous microchannels with Ω-shaped reentrant configurations were developed for application in heat sink cooling. The reentrant porous microchannels were fabricated by using a solid-state sintering method under the replication of specially designed sintering modules. Micro wire electrical discharge machining was utilized to process the graphite-based sintering modules. Two types of commonly used copper powder in heat transfer devices, i.e., spherical and irregular powder, with three fractions of particle sizes respectively, were utilized to construct the porous microchannel heat sinks. The effects of powder type and size on the flow boiling performance of reentrant porous microchannels, i.e., two-phase heat transfer, pressure drop and flow instabilities, were examined under boiling deionized water conditions. The test results show that enhanced two-phase heat transfer was achieved with the increase of particle size for the reentrant porous microchannels with spherical powder, while the reversed trend existed for the counterparts with irregular powder. The reentrant porous microchannels with irregular powder of the smallest particle size presented the best heat transfer performance and lowest pressure drop.

  8. Research progresses and future directions on pool boiling heat transfer

    Directory of Open Access Journals (Sweden)

    M. Kumar

    2015-12-01

    Full Text Available This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated surface. It occurs during boiling of liquids for excess temperature ranging from 5 to 30 °C in various processes related to high vaporization of liquid for specific purposes like sugarcane juice heating for jaggery making, milk heating for khoa making, steam generation, cooling of electronic equipments, refrigeration and etcetera. In this review paper, pool boiling method during heating of liquids for specific purpose is depicted. It is inferred that enhancement in pool boiling heat transfer is a challenging and complex task. Also, recent research and use of various correlations for natural convection pool boiling is reviewed.

  9. High-burn up 10 x 10 100%MOX ABWR core physics analysis with APOLLO2.8 and TRIPOLI-4.5 codes

    Energy Technology Data Exchange (ETDEWEB)

    Blaise, Patrick, E-mail: patrick.blaise@cea.f [Centre de Cadarache, DEN-CAD/DER/SPRC - building 230, F-13108 Saint Paul-Lez-Durance (France); Huot, Nicolas [Centre de Saclay, DEN-DANS/DM2S/SERMA - building 470, F-91191 Gif-sur-Yvette (France); Thiollay, Nicolas [Centre de Cadarache, DEN-CAD/DER/SPEX - building 238, F-13108 Saint Paul-Lez-Durance (France); Fougeras, Philippe; Santamarina, Alain [Centre de Cadarache, DEN-CAD/DER/SPRC - building 230, F-13108 Saint Paul-Lez-Durance (France)

    2010-07-15

    Within the frame of several extensive experimental core physics programs led between 1996 and 2008 between CEA and Japan Nuclear Energy Safety Organization (JNES), the FUBILA experiment has been conducted in the French EOLE Facility between 2005 and 2006 to obtain valuable data for the validation of core analysis methods related to full MOX advanced BWR and high-burn up BWR cores. During this experimental campaign, a particular FUBILA 10 x 10 Advanced BWR configuration devoted to the validation of high-burn up 100%MOX BWR bundles was built. It is characterized by an assembly average total Pu enrichment of 10.6 wt.% and in-channel void of 40%, representative of hot full power conditions at core mid-plane and average discharge burnup of 65 GWd/t. This paper details the validation work led on the TRIPOLI-4.5 Continuous Energy Monte Carlo code and APOLLO2.8/CEA2005V4 deterministic code package for the interpretation of this 10 x 10 high-burn up configuration. The APOLLO2.8/CEA2005V4 package relies on the deterministic lattice transport code APOLLO2.8 based on the Method of Characteristics (MOC), and its new CEA2005v4 multigroup library based on the latest JEFF-3.1.1 nuclear data file, processed also for the TRIPOLI-4.5 code. The results obtained on critical mass and radial pin-by-pin power distributions are presented. For critical mass, the calculation-to-experiment C-E on the k{sub eff} spreads from 300 pcm for TRIPOLI to 600 pcm for APOLLO2.8 in its Optimized BWR Scheme (OBS) in 26 groups. For pin-by-pin radial power distributions, all codes give acceptable results, with maximum discrepancies on C/E - 1 of the order of 3-4% for very heterogeneous bundles where P{sub max}/P{sub min} reaches 4, 2. These values are within 2 standard deviations of the experimental uncertainty. Those results demonstrate the capability of both codes and schemes to accurately predict Advanced High burnup 100%-MOX BWR key-neutron parameters.

  10. Signal processing techniques for sodium boiling noise detection

    International Nuclear Information System (INIS)

    At the Specialists' Meeting on Sodium Boiling Detection organized by the International Working Group on Fast Reactors (IWGFR) of the International Atomic Energy Agency at Chester in the United Kingdom in 1981 various methods of detecting sodium boiling were reported. But, it was not possible to make a comparative assessment of these methods because the signal condition in each experiment was different from others. That is why participants of this meeting recommended that a benchmark test should be carried out in order to evaluate and compare signal processing methods for boiling detection. Organization of the Co-ordinated Research Programme (CRP) on signal processing techniques for sodium boiling noise detection was also recommended at the 16th meeting of the IWGFR. The CRP on Signal Processing Techniques for Sodium Boiling Noise Detection was set up in 1984. Eight laboratories from six countries have agreed to participate in this CRP. The overall objective of the programme was the development of reliable on-line signal processing techniques which could be used for the detection of sodium boiling in an LMFBR core. During the first stage of the programme a number of existing processing techniques used by different countries have been compared and evaluated. In the course of further work, an algorithm for implementation of this sodium boiling detection system in the nuclear reactor will be developed. It was also considered that the acoustic signal processing techniques developed for boiling detection could well make a useful contribution to other acoustic applications in the reactor. This publication consists of two parts. Part I is the final report of the co-ordinated research programme on signal processing techniques for sodium boiling noise detection. Part II contains two introductory papers and 20 papers presented at four research co-ordination meetings since 1985. A separate abstract was prepared for each of these 22 papers. Refs, figs and tabs

  11. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Manglik, R M; Athavale, A; Kalaikadal, D S; Deodhar, A; Verma, U

    2011-09-02

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  12. Multi-scale Control and Enhancement of Reactor Boiling Heat Flux by Reagents and Nanoparticles

    International Nuclear Information System (INIS)

    The phenomenological characterization of the use of non-invasive and passive techniques to enhance the boiling heat transfer in water has been carried out in this extended study. It provides fundamental enhanced heat transfer data for nucleate boiling and discusses the associated physics with the aim of addressing future and next-generation reactor thermal-hydraulic management. It essentially addresses the hypothesis that in phase-change processes during boiling, the primary mechanisms can be related to the liquid-vapor interfacial tension and surface wetting at the solidliquid interface. These interfacial characteristics can be significantly altered and decoupled by introducing small quantities of additives in water, such as surface-active polymers, surfactants, and nanoparticles. The changes are fundamentally caused at a molecular-scale by the relative bulk molecular dynamics and adsorption-desorption of the additive at the liquid-vapor interface, and its physisorption and electrokinetics at the liquid-solid interface. At the micro-scale, the transient transport mechanisms at the solid-liquid-vapor interface during nucleation and bubblegrowth can be attributed to thin-film spreading, surface-micro-cavity activation, and micro-layer evaporation. Furthermore at the macro-scale, the heat transport is in turn governed by the bubble growth and distribution, macro-layer heat transfer, bubble dynamics (bubble coalescence, collapse, break-up, and translation), and liquid rheology. Some of these behaviors and processes are measured and characterized in this study, the outcomes of which advance the concomitant fundamental physics, as well as provide insights for developing control strategies for the molecular-scale manipulation of interfacial tension and surface wetting in boiling by means of polymeric reagents, surfactants, and other soluble surface-active additives.

  13. Microbiological Effectiveness of Disinfecting Water by Boiling in Rural Guatemala

    OpenAIRE

    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 sa...

  14. Modeling of Heat Exchange with Developed Nucleate Boiling on Tenons

    Directory of Open Access Journals (Sweden)

    A. V. Оvsiannik

    2007-01-01

    Full Text Available The paper proposes a thermal and physical model for heat exchange processes with developed nucleate boiling on the developed surfaces (tenons with various contours of heat transfer surface. Dependences for calculating convective heat exchange factor have been obtained on the basis of modeling representation. Investigations have shown that an intensity of convective heat exchange does not depend on tenon profile when boiling takes place on the tenons. The intensity is determined by operating conditions, thermal and physical properties of liquid, internal characteristics of boiling processes and geometrical characteristics of a tenon.

  15. Boiling of Refrigerant R-113. Three-dimensional numerical analysis

    International Nuclear Information System (INIS)

    In this paper a forced convective boiling of Refrigerant R-113 in a vertical annular channel has been simulated by the CFX-5 code. The employed subcooled boiling model uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data. In general a good agreement with the experimental data has been achieved, which shows that the current model may be applied for the Refrigerant R-113 without significantly changing the model parameters. The influence of non-drag forces, bubble diameter size and interfacial drag model on the numerical results has been investigated as well. (author)

  16. Boiling Heat Transfer on Porous Surfaces with Vapor Channels

    Institute of Scientific and Technical Information of China (English)

    吴伟; 杜建华; 王补宣

    2002-01-01

    Boiling heat transfer on porous coated surfaces with vapor channels was investigated experimentally to determine the effects of the size and density of the vapor channels on the boiling heat transfer. Observations showed that bubbles escaping from the channels enhanced the heat transfer. Three regimes were identified: liquid flooding, bubbles in the channel and the bottom drying out region. The maximum heat transfer occurred for an optimum vapor channel density and the boiling heat transfer performance was increased if the channels were open to the bottom of the porous coating.

  17. Subcooled boiling of nano-particle suspensions on Pt wires

    Institute of Scientific and Technical Information of China (English)

    LI Chunhui; WANG Buxuan; PENG Xiaofeng

    2004-01-01

    An experimental investigation is conducted to explore the subcooled boiling characteristics of nano-particle suspensions on Pt wires. Some phenomena are observed for the boiling of water-SiO2 nano-particle suspensions on Pt wires. The experiments show that there exist not any evident differences for boiling of pure water and of nano-particle suspensions at high heat fluxes. However, bubble overlap phenomenon can be easily found for nano-particle suspensions at low heat fluxes, which probably results from the increase of the attracter force between bubbles and of the bubble mass.

  18. Bubble Coalescence Heat Transfer During Subcooled Nucleate Pool Boiling

    Institute of Scientific and Technical Information of China (English)

    Abdoulaye Coulibaly; LIN Xipeng; Bi Jingliang; David M Christopher

    2012-01-01

    Bubble coalescence during subcooled nucleate pool boiling was investigated experimentally using constant wall temperature boundary conditions while the wall heat flux was measured at a various locations to understand the effects of coalescence on the heat transfer. The observations showed that the coalesced bubble moved and oscillated on the heater surface with significant heat transfer variations prior to departure. Some observations also showed coalescence with no increase in the heat transfer rate. The heat flux for boiling with coalescence fluctuated much more than for single bubble boiling due to the vaporization of the liquid layer trapped between the bubbles.

  19. Critical heat flux in subcooled flow boiling

    Science.gov (United States)

    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

  20. Calculations of severe accident progression in the General Electric Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    General Electric is designing a new nuclear power plant: the Simplified Boiling Water Reactor (SBWR). The SBWR is a passive plant in which the core cooling and decay heat removal safety systems are driven by gravity. To model the plant response to severe accidents, MAAP-SBWR, an advanced version of the Modular Accident Analysis Program (MAAP), has been developed. The main feature of the new code is a flexible containment model. The challenges in modeling the SBWR, the code structure and models, and a sample application to the SBWR are discussed

  1. Theory of hydraulic stability of boiling channels

    International Nuclear Information System (INIS)

    A framework of boiling channel stability theory is analyzed. The fundamental equations are those of STABLE code: Three conservation laws of mass, energy and momentum applied to one-dimensional channel, together with Bankoff' slip and Marinelli-Nelson's pressure drop correlation. These equations are analyzed to yield ''Void Equation'', ''Linearized Void Equation'', ''Volume Conservation Law'' and the ''Flow Impedance'' R(s), defined by the dynamic response of pressure drop to the inlet flow. The impedance contains all the information such a stability index, dominant frequency and damping ratio. It is shown that R is a sum of the form R sub(IA) + N sub(F)-1R sub(D) + N sub(R)R sub(R) + N sub(OR), where N's are non-dimensional parameters and R's characteristic impedances determined by three kinds of parameters, N sub(X), N sub(s) and the power distribution parameter. Systematic edition of the characteristic impedances according to the non-dimensional parameters will reduce the need for case-by-case STABLE calculations. Hydraulic stability of BWR channels under constant system pressure, is a phenomenon with three parameters in view of complexity. Furthermore an analysis is conducted to confirm the above stability structure and three typical instabilities are identified. (auth.)

  2. Burnout heat flux in natural flow boiling

    International Nuclear Information System (INIS)

    Twenty runs of experiments were conducted to determine the critical heat flux for natural flow boiling with water flowing upwards through annuli of centrally heated stainless steel tube. The test section has concentric heated tube of 14mm diameter and heated lengthes of 15 and 25 cm. The outside surface of the annulus was formed by various glass tubes of 17.25, 20 and 25.9mm diameter. System pressure is atmospheric. Inlet subcooling varied from 18 to 50C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm2. A number of parameters having dominant influence on the critical heat flux and hydrodynamic instability (flow and pressure oscillations) preceeding the burnout have been studied. These parameters are mass flow rate, mass velocity, throttling, channel geometry (diameters ratio, length to diameter ratio, and test section length), and inlet subcooling. Flow regimes before and at the moments of burnout were observed, discussed, and compared with the existing physical model of burnout

  3. Visualization of boiling flow structure in a natural circulation boiling loop

    Energy Technology Data Exchange (ETDEWEB)

    Karmakar, Arnab; Paruya, Swapan, E-mail: swapanparuya@gmail.com

    2015-04-15

    Highlights: • Vapor–liquid jet flows in natural circulation boiling loop. • Flow patterns and their transitions during geysering instability in the loop. • Evaluation of the efficiency of the needle probe in detecting the vapor–liquid and boiling flow structure. - Abstract: The present study reports vapor–liquid jet flows, flow patterns and their transitions during geysering instability in a natural circulation boiling loop under varied inlet subcooling ΔT{sub sub} (30–50 °C) and heater power Q (4–5 kW). Video imaging, voltage measurement using impedance needle probe, measurement of local pressure and loop flow rate have been carried out in this study. Power spectra of the voltage, the pressure and the flow rate reveal that at a high ΔT{sub sub} the jet flows have long period (21.36–86.95 s) and they are very irregular with a number of harmonics. The period decreases and becomes regular with a decrease of ΔT{sub sub}. The periods of the jet flows at ΔT{sub sub} = 30–50 °C and Q = 4 kW are in close agreement with those obtained from the video imaging. The probe was found to be more efficient than the pressure sensor in detecting the jet flows within an uncertainty of 9.5% and in detecting a variety of bubble classes. Both the imaging and the probe consistently identify the bubbly flow/vapor-mushrooms transition or the bubbly flow/slug flow transition on decreasing ΔT{sub sub} or on increasing Q.

  4. Zero Boil Off System for Cryogen Storage Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This work proposes to develop a zero boil off (ZBO) dewar using a two-stage pulse-tube cooler together with two innovative, continuous-flow cooling loops and an...

  5. Boiling local heat transfer enhancement in minichannels using nanofluids.

    Science.gov (United States)

    Chehade, Ali Ahmad; Gualous, Hasna Louahlia; Le Masson, Stephane; Fardoun, Farouk; Besq, Anthony

    2013-03-18

    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.

  6. Boiling of HFE-7100 on a Straight Pin Fin

    Institute of Scientific and Technical Information of China (English)

    Z. W. Liu; W.W. Lin; D.J. Lee; J.P. Hsu

    2001-01-01

    This paper deals with an experimental investigation of pin fin boiling of saturated and subcooled HFE-7100 under atmospheric pressure. Fin base temperature and heat flux data are measured along with the fin tip temperature. The basic features of boiling stability of HFE-7100 boiling on pin fin had been reported for the first time. For a given liquid/heating surface combination there exist upper steady-state (USS) branch and lower steady-state (LSS)branch, and a large, unstable regime located in between. Zones with different stability characteristics are mapped according to boiling on fins with different aspect ratios. Liquid subcooling can largely enhance heat transfer performance. A longer fin can provide a safer operation.

  7. Boiling and burnout phenomena under transient heat input, 1

    International Nuclear Information System (INIS)

    In order to simulate the thermo-hydrodynamic conditions at reactor power excursions, a test piece was placed in a forced convective channel and heated with exponential power inputs. The boiling heat transfer and the burnout heat flux under the transient heat input were measured, and pressure and water temperature changes in the test section were recorded at the same time. Following experimental results were obtained; (1) Transient boiling heat transfer characteristics at high heat flux stayed on the stationary nucleate boiling curve of each flow condition, or extrapolated line of the curves. (2) Transient burnout heat flux increased remarkably with decreasing heating-time-constant, when the flow rate was lower and the subcooling was higher. (3) Transient burnout phenomena were expressed with the relation of (q sub(max) - q sub(sBO)) tau = constant at several flow conditions. This relation was derived from the stationary burnout mechanism of pool boiling. (auth.)

  8. Deterministic modeling of 100%MOX ABWR lattices with increasing void fraction. Validation of the REL2005 code package on FUBILA experimental program

    International Nuclear Information System (INIS)

    This paper details the validation work led on the REL2005 code package, applied to the modeling of 100%MOX 9x9 Advanced BWR assemblies with increasing void fraction (0 to 70%). The REL2005 package relies on the deterministic lattice transport code APOLLO2.8 based on the Method of Characteristics (MOC), and its new CEA2005 multigroup library based on the latest JEFF-3.1.1 nuclear data file. We describe the overall results obtained on 3 critical cores of the FUBILA experimental program that took place in the EOLE facility between 2005 and 2006: FUBILA/REF (0% void), NORM (40% void) and 70%VOID, for critical masses, void reactivity coefficients and pin-by-pin power distributions by using the REL2005-BWR optimized scheme (26 energy groups coupled with unstructured 2D geometry). For critical masses, the calculation-to-experimentratio C/E on the keff increases with the void fraction from 200 pcm for the REF core to 500 pcm at 70%void. A reference full 3D calculation made with TRIPOLI-4.5 Monte Carlo confirms this trend, as for the void coefficient. For pin-by-pin radial power distributions, REL2005 gives acceptable results, with maximum discrepancies of the order of 3 to 4%, as void fraction increases. These values are within 2 standard deviations of the experimental uncertainty. Those results lead us to be confident in the capability of the REL2005 code package to accurately predict 100%MOX BWR key-neutron parameters. (author)

  9. 40 CFR 180.1056 - Boiled linseed oil; exemption from requirement of tolerance.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Boiled linseed oil; exemption from... From Tolerances § 180.1056 Boiled linseed oil; exemption from requirement of tolerance. Boiled linseed... “boiled linseed oil.” This exemption is limited to use on rice before edible parts form....

  10. An Analytical Approach for Relating Boiling Points of Monofunctional Organic Compounds to Intermolecular Forces

    Science.gov (United States)

    Struyf, Jef

    2011-01-01

    The boiling point of a monofunctional organic compound is expressed as the sum of two parts: a contribution to the boiling point due to the R group and a contribution due to the functional group. The boiling point in absolute temperature of the corresponding RH hydrocarbon is chosen for the contribution to the boiling point of the R group and is a…

  11. Characteristics of phenomenon and sound in microbubble emission boiling

    International Nuclear Information System (INIS)

    Background: Nowadays, the efficient heat transfer technology is required in nuclear energy. Therefore, micro-bubble emission boiling (MEB) is getting more attentions from many researchers due to its extremely high heat-transfer dissipation capability. Purpose: An experimental setup was built up to study the correspondences between the characteristics on the amplitude spectrum of boiling sound in different boiling modes. Methods: The heat element was a copper block heated by four Si-C heaters. The upper of the copper block was a cylinder with the diameter of 10 mm and height of 10 mm. Temperature data were measured by three T-type sheathed thermocouples fitted on the upper of the copper block and recorded by NI acquisition system. The temperature of the heating surface was estimated by extrapolating the temperature distribution. Boiling sound data were acquired by hydrophone and processed by Fourier transform. Bubble behaviors were captured by high-speed video camera with light system. Results: In nucleate boiling region, the boiling was not intensive and as a result, the spectra didn't present any peak. While the MEB fully developed on the heating surface, an obvious peak came into being around the frequency of 300 Hz. This could be explained by analyzing the video data. The periodic expansion and collapse into many extremely small bubbles of the vapor film lead to MEB presenting an obvious characteristic peak in its amplitude spectrum. Conclusion: The boiling mode can be distinguished by its amplitude spectrum. When the MEB fully developed, it presented a characteristic peak in its amplitude spectrum around the frequency between 300-400 Hz. This proved that boiling sound of MEB has a close relation with the behavior of vapor film. (authors)

  12. Numerical Simulation of Pool Boiling from Reentrant Type Structured Surfaces

    OpenAIRE

    Dietl, Jochen

    2015-01-01

    Enhancement of heat transfer in pool boiling can be achieved by employing a structured surface. So called reentrant type surfaces, consisting of subsurface tunnels connected through pores with the pool, were found to strongly improve the performance of heat exchanger tubes. Although employed since decades, several of the processes within the tunnel are not understood and the presented models are not able to predict the different boiling modes. With the rapid development of numerical method...

  13. Hysteresis of boiling for different tunnel-pore surfaces

    OpenAIRE

    Pastuszko Robert; Piasecka Magdalena

    2015-01-01

    Analysis of boiling hysteresis on structured surfaces covered with perforated foil is proposed. Hysteresis is an adverse phenomenon, preventing high heat flux systems from thermal stabilization, characterized by a boiling curve variation at an increase and decrease of heat flux density. Experimental data were discussed for three kinds of enhanced surfaces: tunnel structures (TS), narrow tunnel structures (NTS) and mini-fins covered with the copper wire net (NTS-L). The experiments were carrie...

  14. Heat transfer and pressure drop in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

    This Brief addresses the phenomena of heat transfer and pressure drop in flow boiling in micro channels occurring in high heat flux electronic cooling. A companion edition in the Springer Brief Subseries on Thermal Engineering and Applied Science to “Critical Heat Flux in Flow Boiling in Micro channels,” by the same author team, this volume is idea for professionals, researchers and graduate students concerned with electronic cooling.

  15. Development of an experimental apparatus for nucleate boiling analysis

    International Nuclear Information System (INIS)

    An experimental apparatus is developed for the study of the parameters that affect nucleate boiling. The experimental set up is tested for nucleate boiling in an annular test section with subcooled water flow. The following parameters are analysed: pressure, fluid velocity and the fluid temperature at the test section entrance. The performance of the experimental apparatus is analysed by the results and by the problems raised by the operation of the setup. (Author)

  16. Coupling of wall boiling with discrete population balance model

    International Nuclear Information System (INIS)

    A coupling between a polydisperse population balance method (Multiple Size Group Model - MUSIG) and the RPI wall boiling model for nucleate subcooled boiling has been implemented in ANSYS CFX. It allows more accurate prediction of the interfacial area density for mass, momentum and energy transfer between phases in comparison to the usual local-monodisperse bubble size assumption and underlying bulk bubble diameter correlations as they are commonly used in boiling flow applications like e.g. the prediction of subcooled nucleate boiling in rod bundles and fuel assemblies of PWR. The paper outlines the methodology of the coupled CFD model, which automatically avoids possible inconsistencies in the model formulation for the heated wall, when the generated steam bubbles on the heater surface are injected exactly in the bubble size class corresponding to the predicted bubble departure diameter. The coupling of the RPI wall boiling model and the MUSIG model has been implemented for both homogenous/inhomogeneous variants of the MUSIG model. The paper presents the validation of the coupled modeling approach for the well known test case of nucleate subcooled boiling of R113 refrigerant in a circular annulus with inner heated rod based on the experiments of Roy et al. ANSYS CFX results with the newly implemented approach as well as comparison to data and locally-monodisperse simulations are provided. (author)

  17. Void fraction prediction in saturated flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Francisco J Collado [Dpto de Ingenieria Mecanica-Motores Termicos, CPS-B, Universidad de Zaragoza, Maria de Luna 50018-Zaragoza (Spain)

    2005-07-01

    Full text of publication follows: An essential element in thermal-hydraulics is the accurate prediction of the vapor void fraction, or fraction of the flow cross-sectional area occupied by steam. Recently, the author has suggested to calculate void fraction working exclusively with thermodynamic properties. It is well known that the usual 'flow' quality, merely a mass flow rate ratio, is not at all a thermodynamic property because its expression in function of thermodynamic properties includes the slip ratio, which is a parameter of the process not a function of state. By the other hand, in the classic and well known expression of the void fraction - in function of the true mass fraction of vapor (also called 'static' quality), and the vapor and liquid densities - does not appear the slip ratio. Of course, this would suggest a direct procedure for calculating the void fraction, provided we had an accurate value of the true mass fraction of vapor, clearly from the heat balance. However the classic heat balance is usually stated in function of the 'flow' quality, what sounds really contradictory because this parameter, as we have noted above, is not at all a thermodynamic property. Then we should check against real data the actual relationship between the thermodynamic properties and the applied heat. For saturated flow boiling just from the inlet of the heated tube, and not having into account the kinetic and potential terms, the uniform applied heat per unit mass of inlet water and per unit length (in short, specific linear heat) should be closely related to a (constant) slope of the mixture enthalpy. In this work, we have checked the relation between the specific linear heat and the thermodynamic enthalpy of the liquid-vapor mixture using the actual mass fraction. This true mass fraction is calculated using the accurate measurements of the outlet void fraction taken during the Cambridge project by Knights and Thom in the sixties for

  18. Evaluation of silica behavior for reducing the precoating frequency of the reactor water cleanup system of the Hamaoka NPS

    International Nuclear Information System (INIS)

    When the silica concentration in reactor water exceeds the reference value, the ion-exchange resin powder used in the reactor water cleanup system (CUW) is replaced with new resin powder. This is referred to as the 'precoating of CUW'. Precoating of CUW generates radioactive waste; therefore, a higher frequency of CUW precoating increases the waste disposal cost. In the Advanced Boiling-Water Reactor (ABWR) of Hamaoka Unit 5, the frequency of CUW precoating has been higher than that of any other plants as a result of the high silica concentration in the reactor water; therefore, the behavior of silica in the reactor primary water circuit was examined in order to reduce the frequency of CUW precoating. A calculation model was developed for the silica behavior in the primary water (Silica Behavior Code) and the mass balances of silica in Hamaoka Units 3 and 4 (BWR5), and Unit 5 was then analyzed applying this code. A comparison of these mass balances shows two results for the difference between BWR5 and the ABWR. First, the amount of silica removed from the condensate demineralizer (CD) in the ABWR is less than that in BWR5, because silica, which is transported from the reactor water into the main steam, bypasses the CD and returns to the reactor water directly due to the high pressure heater drain line specific to the ABWR. Second, the amount of silica generated during plant operation in the ABWR is greater than that in BWR5 due to the generation of silica in the high pressure heater drain line. From the above results, it is concluded that the high pressure heater drain line, which is specific to the ABWR, is the cause of the high silica concentration in reactor water in Hamaoka Unit 5. (author)

  19. A phenomenological model of the thermal hydraulics of convective boiling during the quenching of hot rod bundles

    International Nuclear Information System (INIS)

    In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing near-wall effects. Each contribution was considered separately in each of the inverted annular flow (IAF) regimes. The interfacial heat transfer was also formulated as flow-regime dependent. The interfacial drag coefficient model upstream of the CHF point was considered to be similar to flow through a roughened pipe. A free-stream contribution was calculated using Ishii's bubbly flow model for either fully developed subcooled or saturated nucleate boiling. For the drag in the smooth IAF region, a simple smooth-tube correlation for the interfacial friction factor was used. The drag coefficient for the rough-wavy IAF was formulated in the same way as for the smooth IAF model except that the roughness parameter was assumed to be proportional to liquid droplet diameter entrained from the wavy interface. The drag coefficient in the highly dispersed flow regime considered the combined effects of the liquid droplets within the channel and a liquid film on wet unheated walls. 431 refs., 6 figs., 4 tabs

  20. Gravity and Heater Size Effects on Pool Boiling Heat Transfer

    Science.gov (United States)

    Kim, Jungho; Raj, Rishi

    2014-01-01

    The current work is based on observations of boiling heat transfer over a continuous range of gravity levels between 0g to 1.8g and varying heater sizes with a fluorinert as the test liquid (FC-72/n-perfluorohexane). Variable gravity pool boiling heat transfer measurements over a wide range of gravity levels were made during parabolic flight campaigns as well as onboard the International Space Station. For large heaters and-or higher gravity conditions, buoyancy dominated boiling and heat transfer results were heater size independent. The power law coefficient for gravity in the heat transfer equation was found to be a function of wall temperature under these conditions. Under low gravity conditions and-or for smaller heaters, surface tension forces dominated and heat transfer results were heater size dependent. A pool boiling regime map differentiating buoyancy and surface tension dominated regimes was developed along with a unified framework that allowed for scaling of pool boiling over a wide range of gravity levels and heater sizes. The scaling laws developed in this study are expected to allow performance quantification of phase change based technologies under variable gravity environments eventually leading to their implementation in space based applications.

  1. Boiling crisis and non-equilibrium drying transition

    CERN Document Server

    Nikolayev, Vadim

    2016-01-01

    Boiling crisis is the rapid formation of the quasi-continuous vapor film between the heater and the liquid when the heat supply exceeds a critical value. We propose a mechanism for the boiling crisis that is based on the spreading of the dry spot under a vapor bubble. The spreading is initiated by the vapor recoil force, a force coming from the liquid evaporation into the bubble. Since the evaporation intensity increases sharply near the triple contact line, the influence of the vapor recoil can be described as a change of the apparent contact angle. Therefore, for the most usual case of complete wetting of the heating surface by the liquid, the boiling crisis can be understood as a drying transition from complete to partial wetting. The state of nucleate boiling, which is boiling in its usual sense, is characterized by a very large rate of heat transfer from the heating surface to the bulk because the superheated liquid is carried away from the heating surface by the departing vapor bubbles. If the heating p...

  2. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 104 to 105. The data was correlated with the equation Nu = 0.015 Reb0.85 Prb0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  3. Enhanced boiling heat transfer in horizontal test bundles

    Energy Technology Data Exchange (ETDEWEB)

    Trewin, R.R.; Jensen, M.K.; Bergles, A.E.

    1994-08-01

    Two-phase flow boiling from bundles of horizontal tubes with smooth and enhanced surfaces has been investigated. Experiments were conducted in pure refrigerant R-113, pure R-11, and mixtures of R-11 and R-113 of approximately 25, 50, and 75% of R-113 by mass. Tests were conducted in two staggered tube bundles consisting of fifteen rows and five columns laid out in equilateral triangular arrays with pitch-to-diameter ratios of 1.17 and 1.5. The enhanced surfaces tested included a knurled surface (Wolverine`s Turbo-B) and a porous surface (Linde`s High Flux). Pool boiling tests were conducted for each surface so that reference values of the heat transfer coefficient could be obtained. Boiling heat transfer experiments in the tube bundles were conducted at pressures of 2 and 6 bar, heat flux values from 5 to 80 kW/m{sup 2}s, and qualities from 0% to 80%, Values of the heat transfer coefficients for the enhanced surfaces were significantly larger than for the smooth tubes and were comparable to the values obtained in pool boiling. It was found that the performance of the enhanced tubes could be predicted using the pool boiling results. The degradation in the smooth tube heat transfer coefficients obtained in fluid mixtures was found to depend on the difference between the molar concentration in the liquid and vapor.

  4. A correlation for nucleate flow boiling in small channels

    Energy Technology Data Exchange (ETDEWEB)

    Tran, T.N. [Texas Tech Univ., Lubbock, TX (United States). Dept. of Mechanical Engineering]|[Argonne National Lab., IL (United States); Wambsganss, M.W. [Argonne National Lab., IL (United States); Chyu, M.C. [Texas Tech Univ., Lubbock, TX (United States). Dept. of Mechanical Engineering; France, D.M. [Univ. of Illinois, Chicago, IL (United States). Dept. of Mechanical Engineering

    1997-08-01

    Compact heat exchangers are becoming more attractive for applications in which energy conservation, space saving, and cost are important considerations. Applications exist in the process industries where phase-change heat transfer realizes more compact designs and improved performance compared to single-phase heat transfer. However, there have been only a few studies in the literature reporting on phase-change heat transfer and two-phase flow in compact heat exchangers, and validated design correlations are lacking. Recent data from experiments on flow boiling of refrigerants in small channels have led researchers to conclude that nucleation is the dominant heat transfer mechanism over a broad range of heat flux and wall superheats. Local heat transfer coefficients and overall two-phase pressure drops were measured for three different refrigerants with circular and non-circular channels in a range of pressures. This data base supports the nucleate boiling mechanism, and it was used to develop a new correlation for heat transfer in nucleate flow boiling. The correlation is based on the Rohsenow boiling model, introducing a confinement number defined by Kew and Cornwell. The new correlation predicts the experimental data for nucleate flow boiling of three refrigerants within {+-}15%.

  5. Experimental Study on the Thermal Stratification in a Pool Boiling with a Horizontal Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Ryu, Sung Uk; Euh, Dong-Jin; Song, Chul-Hwa [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Thermal stratification is formed in horizontal fluid layers with different temperatures, where the warmer fluid layers are situated above the cooler fluid layers. Thermal stratification phenomena are common in pool type reactor systems, such as the liquid-salt cooled advanced high temperature reactor (AHTR) and liquid-metal cooled fast reactor systems such as the sodium fast reactor (SFR). Thermal stratification is increasingly encountered in large pools that are being used as heat sinks in the new generation of advanced reactors. The small-scale pool test was conducted to investigate the thermal stratification phenomena that occurred during the heat-up of a water in a pool. Because turbulence and boiling models affect the natural convection significantly, it is important to obtain local information regarding the fluid velocity and void distribution to determine the relevant physical models. To understand the flow phenomena inside a pool, a non-intrusive technique is adopted to measure the flow velocity field. In this study, the 2D particle image velocimetry (PIV) measurement technique is used to determine the fluid velocity vector field of single- and/or two-phase natural convection flow and thermal stratification in a pool. Detailed velocity measurements using the 2D PIV measurement technique were conducted to investigate single- and/or two-phase natural convection flow and thermal stratification in a pool boiling. In this study, the two-dimensional velocity vector fields as the water temperature increased were experimentally acquired in a pool that contained a horizontal heater rod. The experimental results indicate a large natural convection flow at the region above the heater rod and thermal stratification at the region below the heater rod. The flow of the opposite direction to each other was shown in the region between the heater rod and the thermal boundary layer. This flow pattern will contribute to maintain the thermal stratification and retard the water

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

    Energy Technology Data Exchange (ETDEWEB)

    Vanderhaegen, M. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris (France); Paumel, K. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Seiler, J. M. [CEA, Nuclear Energy Directorate DEN, Laboratory of Physical Chemistry and Multiphase Thermalhydraulics, 38054 Grenoble (France); Tourin, A. [Laboratory of Waves and Acoustics, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris (France); Jeannot, J. P. [CEA, Nuclear Energy Directorate DEN, Laboratory of Instrumentation and Technological Test, Cadarache, 13108 Saint-Paul-lez-Durance (France); Rodriguez, G. [CEA, Nuclear Energy Directorate DEN, Cadarache, 13108 Saint-Paul-lez-Durance (France)

    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 boiling 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)

  7. On the dynamics of bubbles in boiling water

    International Nuclear Information System (INIS)

    Research highlights: → We devote this work to investigate the bubbles dynamics in boiling water. → A simple experiment of laser scattering was designed to obtain dynamical features. → Correlations and non-exponential distributions were found. → A simple model was able to describe several aspects of the system. - Abstract: We investigate the dynamics of many interacting bubbles in boiling water by using a laser scattering experiment. Specifically, we analyze the temporal variations of a laser intensity signal which passed through a sample of boiling water. Our empirical results indicate that the return interval distribution of the laser signal does not follow an exponential distribution; contrariwise, a heavy-tailed distribution has been found. Additionally, we compare the experimental results with those obtained from a minimalist phenomenological model, finding a good agreement.

  8. Pin cooling and dryout in steady local boiling

    International Nuclear Information System (INIS)

    A study is presented of pin cooling and dryout mechanisms in steady local boiling, with the particular objective of understanding the substantial dryout margins observed in the KNS local boiling experiments. Mechanisms for the entry of liquid into the voided region are discussed, and pin cooling by draining liquid films deduced to be likely. The conditions required for interruption of the film flow, and hence for dryout, are examined, with particular attention to vapour/liquid interactions causing film breakdown, inhibition of rewetting and film flooding. This leads to the hypothesis that dryout occurs when a critical vapour velocity is reached, which is shown to be consistent with the limited data on dryout conditions in steady boiling. (orig.)

  9. Changes of enthalpy slope in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco J.; Monne, Carlos [Universidad de Zaragoza-CPS, Departamento de Ingenieria Mecanica-Motores Termicos, Zaragoza (Spain); Pascau, Antonio [Universidad de Zaragoza-CPS, Departamento de Ciencia de los Materiales y Fluidos-Mecanica de Fluidos, Zaragoza (Spain)

    2006-03-01

    Void fraction data in subcooled flow boiling of water at low pressure measured by General Electric in the 1960s are analyzed following the classical model of Griffith et al. (in Proceedings of ASME-AIChE heat transfer conference, 58-HT-19, 1958). In addition, a new proposal for analyzing one-dimensional steady flow boiling is used. This is based on the physical fact that if the two phases have different velocities, they cannot cover the same distance - the control volume length - in the same time. So a slight modification of the heat balance is suggested, i.e., the explicit inclusion of the vapor-liquid velocity ratio or slip ratio as scaling time factor between the phases, which is successfully checked against the data. Finally, the prediction of void fraction using correlations of the net rate of change of vapor enthalpy in the fully developed regime of subcooled flow boiling is explored. (orig.)

  10. Changes of enthalpy slope in subcooled flow boiling

    Science.gov (United States)

    Collado, Francisco J.; Monné, Carlos; Pascau, Antonio

    2006-03-01

    Void fraction data in subcooled flow boiling of water at low pressure measured by General Electric in the 1960s are analyzed following the classical model of Griffith et al. (in Proceedings of ASME-AIChE heat transfer conference, #58-HT-19, 1958). In addition, a new proposal for analyzing one-dimensional steady flow boiling is used. This is based on the physical fact that if the two phases have different velocities, they cannot cover the same distance—the control volume length—in the same time. So a slight modification of the heat balance is suggested, i.e., the explicit inclusion of the vapor liquid velocity ratio or slip ratio as scaling time factor between the phases, which is successfully checked against the data. Finally, the prediction of void fraction using correlations of the net rate of change of vapor enthalpy in the fully developed regime of subcooled flow boiling is explored.

  11. Boiling heat transfer on fins – experimental and numerical procedure

    Directory of Open Access Journals (Sweden)

    Orzechowski T.

    2014-03-01

    Full Text Available The paper presents the research methodology, the test facility and the results of investigations into non-isothermal surfaces in water boiling at atmospheric pressure, together with a discussion of errors. The investigations were conducted for two aluminium samples with technically smooth surfaces and thickness of 4 mm and 10 mm, respectively. For the sample of lower thickness, on the basis of the surface temperature distribution measured with an infrared camera, the local heat flux and the heat transfer coefficient were determined and shown in the form of a boiling curve. For the thicker sample, for which 1-D model cannot be used, numerical calculations were conducted. They resulted in obtaining the values of the local heat flux on the surface the invisible to the infrared, camera i.e. on the side on which the boiling of the medium proceeds.

  12. On mechanism of explosive boiling in nanosecond regime

    Science.gov (United States)

    Çelen, Serap

    2016-06-01

    Today laser-based machining is used to manufacture vital parts for biomedical, aviation and aerospace industries. The aim of the paper is to report theoretical, numerical and experimental investigations of explosive boiling under nanosecond pulsed ytterbium fiber laser irradiation. Experiments were performed in an effective peak power density range between 1397 and 1450 MW/cm2 on pure titanium specimens. The threshold laser fluence for phase explosion, the pressure and temperature at the target surface and the velocity of the expulsed material were reported. A narrow transition zone was realized between the normal vaporization and phase explosion fields. The proof of heterogeneous boiling was given with detailed micrographs. A novel thermal model was proposed for laser-induced splashing at high fluences. Packaging factor and scattering arc radius terms were proposed to state the level of the melt ejection process. Results of the present investigation explain the explosive boiling during high-power laser interaction with metal.

  13. Boiling heat transfer in porous media composed of particles

    International Nuclear Information System (INIS)

    The boiling heat transfer in the porous media composed of spherical fuel elements exerts significant influences on the reactor's efficiency and safety. In the present study an experimental setup was designed and the boiling heat transfer in the porous media composed of spheres of regular distribution was investigated. Four spheres with diameters of 5mm, 6mm, 7mm and 8mm were used in the test sections. The greater particle diameter led to lower heat transfer coefficient, and resulted in higher wall superheat of original nucleation boiling. The variation of heat transfer coefficient was divided into three groups according to two-phase flow patterns and void fraction. A correlation of heat transfer coefficient was proposed with a mean relative deviation of ± 16%. (author)

  14. Visualization of pool boiling from complex surfaces with internal tunnels

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2012-04-01

    Full Text Available The paper presents experimental investigations of boiling heat transfer for a system of connected narrow horizontal and vertical tunnels. These extended surfaces, named narrow tunnel structure (NTS, can be applied to electronic element cooling. The experiments were carried out with ethanol at atmospheric pressure. The tunnel external covers were manufactured out of 0.1 mm thick perforated copper foil (hole diameters 0.5 mm, sintered with the mini-fins, formed on the vertical side of the 10 mm high rectangular fins and horizontal inter-fin surface. Visualization studies were conducted with a transparent structured model of joined narrow tunnels limited with the perforated foil. The visualization investigations aimed to formulate assumptions for the boiling model through distinguishing boiling types and defining all phases of bubble growth.

  15. Boiling and burnout phenomena under transient heat input, 1

    International Nuclear Information System (INIS)

    This paper reports in the experimental results concerning unsteady burnout phenomenon, based on unsteady boiling heat transfer data, burnout heat flux data and the data of changing pressure and water temperature in course of time. These data were acquired by unsteady heating of gas-liquid two phase flow. This experiment simulated the thermohydrodynamic conditions under the runaway power of a nuclear reactor. The following results have been clarified. The boiling with high heat flux showed the same heat transfer characteristics as the steady nuclear boiling curves under each flow condition. Under the conditions of low flow speed and high sub-cool degree, the unsteady burnout heat flux showed the extreme increase of the maximum heat flux owing to the shortening of the time constant. The generation of unsteady burnout phenomena is dominated by two phase flow conditions and by bubble behavior near the heat transfer surface owing to the change of heating conditions and flow conditions. (Tai, I.)

  16. Numerical simulation of pool boiling of a Lennard-Jones liquid

    KAUST Repository

    Inaoka, Hajime

    2013-09-01

    We performed a numerical simulation of pool boiling by a molecular dynamics model. In the simulation, a liquid composed of Lennard-Jones particles in a uniform gravitational field is heated by a heat source at the bottom of the system. The model successfully reproduces the change in regimes of boiling from nucleate boiling to film boiling with the increase of the heat source temperature. We present the pool boiling curve by the model, whose general behavior is consistent with those observed in experiments of pool boiling. © 2013 Elsevier B.V. All rights reserved.

  17. Immersion cooling nucleate boiling of high power computer chips

    International Nuclear Information System (INIS)

    Highlights: ► Experimental investigations of nucleate boiling of dielectric liquids on porous graphite (PG). ► Marked enhancements in nucleate boiling heat transfer coefficient and CHF. ► Critical heat flux (CHF) increases linearly with increased liquid subcooling. ► PG–Cu spreaders for cooling 10 × 10 computer chips remove up to 100 W. - Abstract: This paper presents experimental results of saturation and subcooled boiling of FC-72 and HFE-7100 dielectric liquids on uniformly heated, 10 × 10 mm porous graphite (PG) surfaces for potential applications to immersion cooling of high power computer chips. The experiments investigated the effects of surface inclination, from upward-facing (0°) to downward-facing (180°), and liquid subcooling from 0 to 30 K on nucleate boiling heat transfer coefficient and critical heat flux. The presented experimental data and correlations for natural convection of dielectric liquids on PG and plane surfaces are important for cooling chips while in the standby mode when surface heat flux 2. The experimental curves of the nucleate boiling heat transfer coefficient for FC-72 dielectric liquid in the upward-facing orientation are used in 3-D thermal analysis for sizing and quantifying the performance of copper (Cu), PG and PG–Cu composite spreaders for removing the dissipated thermal power by an underlying 10 × 10 mm computer chip with non-uniform heat dissipation. The 2 mm-thick spreaders are cooled by either saturation or 30 K subcooled nucleate boiling of FC-72 and the composite spreader consists of 0.4 mm-thick surface layer of PG and 1.6 mm-thick Cu substrate.

  18. A high-fidelity approach towards simulation of pool boiling

    Energy Technology Data Exchange (ETDEWEB)

    Yazdani, Miad; Radcliff, Thomas; Soteriou, Marios; Alahyari, Abbas A. [United Technologies Research Center, East Hartford, Connecticut 06108 (United States)

    2016-01-15

    A novel numerical approach is developed to simulate the multiscale problem of pool-boiling phase change. The particular focus is to develop a simulation technique that is capable of predicting the heat transfer and hydrodynamic characteristics of nucleate boiling and the transition to critical heat flux on surfaces of arbitrary shape and roughness distribution addressing a critical need to design enhanced boiling heat transfer surfaces. The macro-scale of the phase change and bubble dynamics is addressed through employing off-the-shelf Computational Fluid Dynamics (CFD) methods for interface tracking and interphase mass and energy transfer. The micro-scale of the microlayer, which forms at early stage of bubble nucleation near the wall, is resolved through asymptotic approximation of the thin-film theory which provides a closed-form solution for the distribution of the micro-layer and its influence on the evaporation process. In addition, the sub-grid surface roughness is represented stochastically through probabilistic density functions and its role in bubble nucleation and growth is then represented based on the thermodynamics of nucleation process. This combination of deterministic CFD, local approximation, and stochastic representation allows the simulation of pool boiling on any surface with known roughness and enhancement characteristics. The numerical model is validated for dynamics and hydrothermal characteristics of a single nucleated bubble on a flat surface against available literature data. In addition, the prediction of pool-boiling heat transfer coefficient is verified against experimental measurements as well as reputable correlations for various roughness distributions and different surface orientations. Finally, the model is employed to demonstrate pool-boiling phenomenon on enhanced structures with reentrance cavities and to explore the effect of enhancement feature design on thermal and hydrodynamic characteristics of these surfaces.

  19. Experimental study of mass boiling in a porous medium model

    International Nuclear Information System (INIS)

    This manuscript presents a pore-scale experimental study of convective boiling heat transfer in a two-dimensional porous medium. The purpose is to deepen the understanding of thermohydraulics of porous media saturated with multiple fluid phases, in order to enhance management of severe accidents in nuclear reactors. Indeed, following a long-lasting failure in the cooling system of a pressurized water reactor (PWR) or a boiling water reactor (BWR) and despite the lowering of the control rods that stops the fission reaction, residual power due to radioactive decay keeps heating up the core. This induces water evaporation, which leads to the drying and degradation of the fuel rods. The resulting hot debris bed, comparable to a porous heat-generating medium, can be cooled down by reflooding, provided a water source is available. This process involves intense boiling mechanisms that must be modelled properly. The experimental study of boiling in porous media presented in this thesis focuses on the influence of different pore-scale boiling regimes on local heat transfer. The experimental setup is a model porous medium made of a bundle of heating cylinders randomly placed between two ceramic plates, one of which is transparent. Each cylinder is a resistance temperature detector (RTD) used to give temperature measurements as well as heat generation. Thermal measurements and high-speed image acquisition allow the effective heat exchanges to be characterized according to the observed local boiling regimes. This provides precious indications precious indications for the type of correlations used in the non-equilibrium macroscopic model used to model reflooding process. (author)

  20. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    Energy Technology Data Exchange (ETDEWEB)

    Jo, HangJin; Kim, Jin Man [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Yeom, Hwasung [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States); Lee, Gi Cheol [Department of Mechanical Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Kiyofumi, Moriyama; Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Sridharan, Kumar; Corradini, Michael [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States)

    2015-09-15

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding.

  1. Hysteresis of boiling for different tunnel-pore surfaces

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2015-01-01

    Full Text Available Analysis of boiling hysteresis on structured surfaces covered with perforated foil is proposed. Hysteresis is an adverse phenomenon, preventing high heat flux systems from thermal stabilization, characterized by a boiling curve variation at an increase and decrease of heat flux density. Experimental data were discussed for three kinds of enhanced surfaces: tunnel structures (TS, narrow tunnel structures (NTS and mini-fins covered with the copper wire net (NTS-L. The experiments were carried out with water, R-123 and FC-72 at atmospheric pressure. A detailed analysis of the measurement results identified several cases of type I, II and III for TS, NTS and NTS-L surfaces.

  2. On Boiling of Crude Oil under Elevated Pressure

    CERN Document Server

    Pimenova, Anastasiya V

    2015-01-01

    We construct a thermodynamic model for theoretical calculation of the boiling process of multicomponent mixtures of hydrocarbons (e.g., crude oil). The model governs kinetics of the mixture composition in the course of the distillation process along with the boiling temperature increase. The model heavily relies on the theory of dilute solutions of gases in liquids. Importantly, our results are applicable for modelling the process under elevated pressure (while the empiric models for oil cracking are not scalable to the case of extreme pressure), such as in an oil field heated by lava intrusions.

  3. On Boiling of Crude Oil under Elevated Pressure

    Science.gov (United States)

    Pimenova, Anastasiya V.; Goldobin, Denis S.

    2016-02-01

    We construct a thermodynamic model for theoretical calculation of the boiling process of multicomponent mixtures of hydrocarbons (e.g., crude oil). The model governs kinetics of the mixture composition in the course of the distillation process along with the boiling temperature increase. The model heavily relies on the theory of dilute solutions of gases in liquids. Importantly, our results are applicable for modelling the process under elevated pressure (while the empiric models for oil cracking are not scalable to the case of extreme pressure), such as in an oil field heated by lava intrusions.

  4. Dimensional analysis of boiling heat transfer burnout conditions

    International Nuclear Information System (INIS)

    The first criteria in boiling water systems design, such as boiling water reactors, is that no burnout in the core is allowed to exist under any conditions of the reactor operation either during steady state operation or during any of the several postulated accidental transients, such as sudden interruption of coolant flow in the reactor core (due to pump failure or blockage of fuel channel). The aim of the present work is to obtain a correlation for the critical heat flux based on a theoretical study where the mechanism of burn out and the related hydrodynamic and heat transfer equations are considered. 8 refs

  5. A Study of the Influence of Solid Particles on Boiling Hysteresis

    Institute of Scientific and Technical Information of China (English)

    M.H.Shi; J.Ma

    1992-01-01

    Experiments have been performed to determine the effects on boiling hysteresis of locally fluidized particles contained in a liquid that serves as coolant for electronic equipment.The results show that Iocally fluidized particles can diminish boiling hysteresis.

  6. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  7. Experiments on microgravity boiling heat transfer by using transparent heaters

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, H. [Kyushu Univ., Fukuoka (Japan). Dept. of Energy and Mech. Eng.

    1997-11-01

    To clarify the relation between the liquid-vapor behavior and the heat transfer characteristics in the boiling phenomena, the structures of transparent heaters were developed for both flow boiling experiments and were applied to the microgravity environment realized by the parabolic flight of aircraft. In the flow boiling experiment, a transparent heated tube makes the heating, the observation of liquid-vapor behavior and the measurement of heat transfer data simultaneously possible. The heat transfer coefficient in the annular flow regime at moderate quality has distinct dependence on gravity provided that the mass velocity is not so high, while no noticeable gravity effect is seen at high quality and in the bubbly flow regime. The measured gravity effect was directly related to the behavior of annular liquid film observed through the transparent tube wall. In the pool boiling experiment, a structure of transparent heating surface realizes both the observation of the macrolayer or microlayer behavior from underneath and the measurements of local surface temperatures and the layer thickness. It was clarified in the microgravity experiments that no vapor stem exists but tiny bubbles are observed in the macrolayer underneath a large coalesced bubble at high heat flux. The heat flux evaluated by the heat conduction across the layer assumes less than 30% of the total to be transferred. The evaporation of the microlayers underneath primary bubbles just after the generation dominates the heat transfer in the microgravity, not only in the isolated bubble region but also in the coalesced bubble region. (orig.) 14 refs.

  8. Electrochemical study of aluminum corrosion in boiling high purity water

    Science.gov (United States)

    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.

  9. STEAM TURBINES WITH A LOW-BOILING WORKING AGENT

    OpenAIRE

    Morozov, N.; Karasev, V.

    2010-01-01

    The subject of the article is the assembly of a steam-generator plant with a natural working agent. A method of calculation for steam turbines with a low-boiling working agent is offered, which accounts for the correlation between the adiabatic curve indication, pressure and temperature in the overheated vapor area.

  10. Corrosion fatigue behavior of zirconium in boiling nitric acid

    International Nuclear Information System (INIS)

    The corrosion fatigue behavior of zirconium in boiling nitric acid has been studied to evaluate the reliability of zirconium used in nuclear fuel reprocessing equipment. An apparatus designed for corrosion fatigue tests in boiling nitric acid was used. The crack growth rate of zirconium was measured as a function of the stress intensity factor using TDCB type specimens. After the tests, the fracture morphology was examined with a scanning electron microscope. The crack growth rate was influenced with the texture of specimens and the test environments. In air at room temperature, the crack growth rate at the longitudinal direction of specimens was faster than that of the transverse direction. Moreover, the crack growth rate in boiling nitric acid was more faster than that in air at room temperature. According to the fractographic examination, X-ray analysis, and so on, the observed results were interpreted with based on the crystal anisotropy on mechanical properties and the susceptibility to stress corrosion cracking in boiling nitric acid of zirconium. (author)

  11. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.)

  12. How long does it take to boil an egg? Revisited

    Energy Technology Data Exchange (ETDEWEB)

    Buay, D [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Foong, S K [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Kiang, D [Department of Physics, Chinese University of Hong Kong, Shatin, New Territories, Hong Kong (China); Kuppan, L [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Centre for Research in Pedagogy and Practice, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Liew, V H [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore)

    2006-01-01

    How long does it take to boil an egg? Theoretical prediction, based on a simple adaptation of the solution to the exact thermal diffusion equation for a sphere, is consistent with experiments. The experimental data are also used to estimate an average value for the thermal diffusivity of an egg.

  13. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    Directory of Open Access Journals (Sweden)

    D. S. Obukhov

    2006-01-01

    Full Text Available The paper contains analysis of heat exchange investigations while boiling hydrocarbon fuel. The obtained data are within the limits of the S.S. Kutateladze dependence proposed in 1939. Heat exchange at non-stationary heat release has not been investigated. The data for hydrocarbon fuel with respect to critical density of heat flow are not available even for stationary conditions.

  14. BORATING OF CARBON AND ALLOY STEEL IN BOILING LAYER

    Directory of Open Access Journals (Sweden)

    N. Koukhareva

    2012-01-01

    Full Text Available The paper describes how to obtain boride coatings on steel 20, 4X5MФС, X12M being treated in a boiling layer of metallothermic powder environment. Phase and chemical compositions, hardness and wear- resistance of boride coatings

  15. Experimental demonstration of contaminant removal from fractured rock by boiling.

    Science.gov (United States)

    Chen, Fei; Liu, Xiaoling; Falta, Ronald W; Murdoch, Lawrence C

    2010-08-15

    This study was conducted to experimentally demonstrate removal of a chlorinated volatile organic compound from fractured rock by boiling. A Berea sandstone core was contaminated by injecting water containing dissolved 1,2-DCA (253 mg/L) and sodium bromide (144 mg/L). During heating, the core was sealed except for one end, which was open to the atmosphere to simulate an open fracture. A temperature gradient toward the outlet was observed when boiling occurred in the core. This indicates that steam was generated and a pressure gradient developed toward the outlet, pushing steam vapor and liquid water toward the outlet. As boiling occurred, the concentration of 1,2-DCA in the condensed effluent peaked up to 6.1 times higher than the injected concentration. When 38% of the pore volume of condensate was produced, essentially 100% of the 1,2-DCA was recovered. Nonvolatile bromide concentration in the condensate was used as an indicator of the produced steam quality (vapor mass fraction) because it can only be removed as a solute, and not as a vapor. A higher produced steam quality corresponds to more concentrated 1,2-DCA removal from the core, demonstrating that the chlorinated volatile compound is primarily removed by partitioning into vapor phase flow. This study has experimentally demonstrated that boiling is an effective mechanism for CVOC removal from the rock matrix.

  16. Computations of film boiling. Part II: multi-mode film boiling

    Energy Technology Data Exchange (ETDEWEB)

    Esmaeeli, A.; Tryggvason, G. [Worcester Polytechnic Institute, MA (United States). Mechanical Engineering Department

    2004-12-01

    Film boiling on horizontal periodic surfaces is investigated by direct numerical simulations. A front tracking/finite difference technique is used to solve the momentum and the energy equations in both phases and to account for inertia, viscosity, and surface deformation. Effect of the unit cell size W on the interface dynamics, heat transfer, and fluid flow is studied for different wall superheats. The simulations are carried out over sufficiently long times to capture several bubble release cycles ands to evaluate the quasi steady-state Nusselt number (Nu). While instantaneous Nusselt number will change as result of a change in the system size, statistically steady-state Nusselt number remains almost the same. Simulations of two-dimensional systems in large unit cells, 5{lambda}{sub d2} < W < 10{lambda}{sub d2}, show a distribution of bubble spacing in the range of 0.61{lambda}{sub d2}-1.46{lambda}{sub d2}. At relatively low superheats (Ja {<=} 0.064) the bubbles are released periodically from the vapor film, but at intermediate superheats (0.064 < Ja < 2.13) permanent vapor jets are formed with no bubble break off. At sufficiently high superheats, the vapor jets start to interact. It is shown that the average bubble spacing does not change with changes in the wall superheat. (author)

  17. 46 CFR 154.705 - Cargo boil-off as fuel: General.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: General. 154.705 Section 154.705... Pressure and Temperature Control § 154.705 Cargo boil-off as fuel: General. (a) Each cargo boil-off fuel system under § 154.703(c) must meet §§ 154.706 through 154.709. (b) The piping in the cargo boil-off...

  18. Determination of Boiling Range of Xylene Mixed in PX Device Using Artificial Neural Networks

    OpenAIRE

    Zhu, Ting; Zhu, Yuxuan; Yang, Hong; Li, Hao

    2014-01-01

    Determination of boiling range of xylene mixed in PX device is currently a crucial topic in the practical applications because of the recent disputes of PX project in China. In our study, instead of determining the boiling range of xylene mixed by traditional approach in laboratory or industry, we successfully established two Artificial Neural Networks (ANNs) models to determine the initial boiling point and final boiling point respectively. Results show that the Multilayer Feedforward Neural...

  19. Experimental Evidence of the Vapor Recoil Mechanism in the Boiling Crisis

    OpenAIRE

    Nikolayev, Vadim; Chatain, D.; Garrabos, Y.; Beysens, D.

    2006-01-01

    International audience Boiling crisis experiments are carried out in the vicinity of the liquid-gas critical point of H2. A magnetic gravity compensation setup is used to enable nucleate boiling at near critical pressure. The measurements of the critical heat flux that defines the threshold for the boiling crisis are carried out as a function of the distance from the critical point. The obtained power law behavior and the boiling crisis dynamics agree with the predictions of the vapor reco...

  20. Some specific features of subcooled boiling heat transfer and crisis at extremely high heat flux densities

    Energy Technology Data Exchange (ETDEWEB)

    Gotovsky, M.A. [Polzunov Institute, Saint Petersburg (Russian Federation)

    2001-07-01

    Forced convection boiling is the process used widely in a lot of industry branches including NPP. Heat transfer intensity under forced convection boiling is considered in different way in dependence on conditions. One of main problems for the process considered is an influence of interaction between forced flow and boiling on heat transfer character. For saturated water case a transition from ''pure'' forced convection to nucleate boiling can be realized in smooth form. (author)

  1. Numerical Investigation of Microgravity Tank Pressure Rise Due to Boiling

    Science.gov (United States)

    Hylton, Sonya; Ibrahim, Mounir; Kartuzova, Olga; Kassemi, Mohammad

    2015-01-01

    The ability to control self-pressurization in cryogenic storage tanks is essential for NASAs long-term space exploration missions. Predictions of the tank pressure rise in Space are needed in order to inform the microgravity design and optimization process. Due to the fact that natural convection is very weak in microgravity, heat leaks into the tank can create superheated regions in the liquid. The superheated regions can instigate microgravity boiling, giving rise to pressure spikes during self-pressurization. In this work, a CFD model is developed to predict the magnitude and duration of the microgravity pressure spikes. The model uses the Schrage equation to calculate the mass transfer, with a different accommodation coefficient for evaporation at the interface, condensation at the interface, and boiling in the bulk liquid. The implicit VOF model was used to account for the moving interface, with bounded second order time discretization. Validation of the models predictions was carried out using microgravity data from the Tank Pressure Control Experiment, which flew aboard the Space Shuttle Mission STS-52. Although this experiment was meant to study pressurization and pressure control, it underwent boiling during several tests. The pressure rise predicted by the CFD model compared well with the experimental data. The ZBOT microgravity experiment is scheduled to fly on February 2016 aboard the ISS. The CFD model was also used to perform simulations for setting parametric limits for the Zero-Boil-Off Tank (ZBOT) Experiments Test Matrix in an attempt to avoid boiling in the majority of the test runs that are aimed to study pressure increase rates during self-pressurization. *Supported in part by NASA ISS Physical Sciences Research Program, NASA HQ, USA

  2. Modeling acid-gas generation from boiling chloride brines

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guoxiang; Spycher, Nicolas; Sonnenthal, Eric; Steefel, Carl

    2009-11-16

    This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150 C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC) processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual and modeled distillation experiments do not represent

  3. Modeling acid-gas generation from boiling chloride brines

    Directory of Open Access Journals (Sweden)

    Sonnenthal Eric

    2009-11-01

    Full Text Available Abstract Background This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Results Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150°C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. Conclusion The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual

  4. Bubble and boundary layer behaviour in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Maurus, Reinhold; Sattelmayer, Thomas [Lehrstuhl fuer Thermodynamik, Technische Universitaet Muenchen, 85747 Garching (Germany)

    2006-03-15

    Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. In the study, an experimental investigation in the nucleate boiling regime was performed for water circulating in a closed loop at atmospheric pressure. The horizontal orientated test-section consists of a rectangular channel with a one side heated copper strip and good optical access. Various optical observation techniques were applied to study the bubble behaviour and the characteristics of the fluid phase. The bubble behaviour was recorded by the high-speed cinematography and by a digital high resolution camera. Automated image processing and analysis algorithms developed by the authors were applied for a wide range of mass flow rates and heat fluxes in order to extract characteristic length and time scales of the bubbly layer during the boiling process. Using this methodology, the bubbles were automatically analysed and the bubble size, bubble lifetime, waiting time between two cycles were evaluated. Due to the huge number of observed bubbles a statistical analysis was performed and distribution functions were derived. Using a two-dimensional cross-correlation algorithm, the averaged axial phase boundary velocity profile could be extracted. In addition, the fluid phase velocity profile was characterised by means of the particle image velocimetry (PIV) for the single phase flow as well as under subcooled flow boiling conditions. The results indicate that the bubbles increase the flow resistance. The impact on the flow exceeds by far the bubbly region and it depends on the magnitude of the boiling activity. Finally, the ratio of the averaged phase boundary velocity and of the averaged fluid velocity was evaluated for the bubbly region. (authors)

  5. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  6. Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ali, Amir [Univ. of New Mexico, Albuquerque, NM (United States); Liu, Maolong [Univ. of New Mexico, Albuquerque, NM (United States); Blandford, Edward [Univ. of New Mexico, Albuquerque, NM (United States)

    2016-06-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (i.e., contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.

  7. MRI monitoring of lesions created at temperature below the boiling point and of lesions created above the boiling point using high intensity focused ultrasound

    OpenAIRE

    Damianou, C.; Ioannides, K.; Hadjisavvas, V.; Mylonas, N.; Couppis, A.; Iosif, D.; Kyriacou, P. A.

    2010-01-01

    Magnetic Resonance Imaging (MRI) was utilized to monitor lesions created at temperature below the boiling point and lesions created at temperature above the boiling point using High Intensity Focused Ultrasound (HIFU) in freshly excised kidney, liver and brain and in vivo rabbit kidney and brain. T2-weighted fast spin echo (FSE) was proven as an excellent MRI sequence that can detect lesions with temperature above the boiling point in kidney. This advantage is attributed to the significant di...

  8. TRAC-BD1: transient reactor analysis code for boiling-water systems

    Energy Technology Data Exchange (ETDEWEB)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented.

  9. Experimental investigation into the effects of coolant additives on boiling phenomena in pressurized water reactors

    International Nuclear Information System (INIS)

    This study investigates the effects of coolant additives like boric acid on boiling phenomena in pressurized water reactors under conditions as realistic as possible. The effects covered range from subcooled boiling to critical boiling conditions (CHF). The focus of this project lies on flow boiling with up to 40 bar and 250 °C in order to generate a data basis for a possible extrapolation to reactor conditions. The results of the experiments are used to implement and validate new models into CFD-Codes in context to a nationwide German joint research project with the specific aim of improving CFD boiling-models. (author)

  10. A New Theory of Nucleate Pool Boiling in Arbitrary Gravity

    Science.gov (United States)

    Buyevich, Y. A.; Webbon, Bruce W.

    1995-01-01

    Heat transfer rates specific to nucleate pool boiling under various conditions are determined by the dynamics of vapour bubbles that are originated and grow at nucleation sites of a superheated surface. A new dynamic theory of these bubbles has been recently developed on the basis of the thermodynamics of irreversible processes. In contrast to other existing models based on empirically postulated equations for bubble growth and motion, this theory does not contain unwarrantable assumptions, and both the equations are rigorously derived within the framework of a unified approach. The conclusions of the theory are drastically different from those of the conventional models. The bubbles are shown to detach themselves under combined action of buoyancy and a surface tension force that is proven to add to buoyancy in bubble detachment, but not the other way round as is commonly presumed. The theory ensures a sound understanding of a number of so far unexplained phenomena, such as effect caused by gravity level and surface tension on the bubble growth rate and dependence of the bubble characteristics at detachment on the liquid thermophysical parameters and relevant temperature differences. The theoretical predictions are shown to be in a satisfactory qualitative and quantitative agreement with observations. When being applied to heat transfer at nucleate pool boiling, this bubble dynamic theory offers an opportunity to considerably improve the main formulae that are generally used to correlate experimental findings and to design boiling heat removal in various industrial applications. Moreover, the theory makes possible to pose and study a great deal of new problems of essential impact in practice. Two such problems are considered in detail. One problem concerns the development of a principally novel physical model for the first crisis of boiling. This model allows for evaluating critical boiling heat fluxes under various conditions, and in particular at different

  11. Numerical modeling of boiling heat transfer in porous media

    International Nuclear Information System (INIS)

    Theoretical models were developed and validated to investigate boiling heat transfer in porous layers with and without the presence of chimneys. The critical heat flux and distributions of temperature, liquid saturation, liquid and vapor pressures, and liquid and vapor velocities were predicted numerically under typical PWR conditions. The results indicate that a porous layer produces a higher heat transfer coefficient in the nucleate boiling regime, as is well-known, and could potentially yield a much higher critical heat flux than a plain surface does. Moreover, a chimney-type porous layer can have a better thermal performance, i.e., heat transfer coefficient and critical heat flux than a homogeneous one, primarily due to the presence of chimneys providing pathways for vapor to escape from the porous layer with less resistance

  12. Catastrophe characteristics of the condensation and pool boiling phenomena

    Science.gov (United States)

    Ma, Xuehu; Xu, Dunqi; Lin, Jifang

    1995-02-01

    Recently, Utaka proposed two types of the transition modes of dropwise condensation, i.e. the continuous and the jumping modes, and presented a criterion for determining the condensation transition mode. Stylianous and Rose proposed two hypotheses, the coalescence-limited transition and the nucleation site saturation transition. Neither Utaka's criterion nor Rose's hypotheses could clearly interpret the physical mechanisms of the transition both from filmwise to dropwise and from dropwise to pseudofilm condensation, and explicitly presented the main factors affecting the transitions. Kalinin hs given a general review of the transition boiling heat transfer. The catastrophe theory will be applied here to eluicidate the complex phenomena of the transitions of the condensation and boiling pattern states.

  13. Computations of film boiling. Part I: numerical method

    Energy Technology Data Exchange (ETDEWEB)

    Esmaeeli, A.; Tryggvason, G. [Worcester Polytechnic Institute, MA (United States). Mechanical Engineering Department

    2004-12-01

    A numerical method for direct simulations of boiling flows is presented. The method is similar to the front tracking/finite difference technique of Juric and Tryggvason [Int. J. Multiphase Flow 24 (1998) 387], where one set of conservation equations is used to represent the mass transfer, heat transfer, and fluid flow in the liquid and the vapor, but improves on their numerical technique by elimination of their iterative algorithm. The justification of the mathematical formulation is presented and the numerical method and the code is validated by comparison of the results with the exact solutions of a few analytical problems. A grid refinement test for film boiling on a horizontal surface shows the convergence of results. (author)

  14. Problem of Boil - off in LNG Supply Chain

    OpenAIRE

    Dobrota, Đorđe; Lalić, Branko; Komar, Ivan

    2013-01-01

    This paper examines the problem of evaporation of Liquefied Natural Gas (LNG) occurring at different places in the LNG supply chain. Evaporation losses in the LNG supply chain are one of the key factors for LNG safety, technical and economic assessment. LNG is stored and transported in tanks as a cryogenic liquid, i.e. as a liquid at a temperature below its boiling point at near atmospheric pressure. Due to heat entering the cryogenic tank during storage and transportatio...

  15. Evaluation of boiled potato peel as a wound dressing.

    Science.gov (United States)

    Dattatreya, R M; Nuijen, S; van Swaaij, A C; Klopper, P J

    1991-08-01

    In a series of experiments full thickness skin defects in 68 rats were covered with dressings made of boiled potato peels according to the method developed in Bombay. The wounds closed within 14 days and histologically complete repair of epidermis was found. The cork layer of the potato peel prevents dehydration of the wound and protects against exogenous agents. Experiments with homogenates revealed that a complete structure of the peel is necessary. Steroidal glycosides may have contributed to the favourable results. PMID:1930669

  16. Phase field model for the study of boiling

    International Nuclear Information System (INIS)

    This study concerns both the modeling and the numerical simulation of boiling flows. First we propose a review concerning nucleate boiling at high wall heat flux and focus more particularly on the current understanding of the boiling crisis. From this analysis we deduce a motivation for the numerical simulation of bubble growth dynamics. The main and remaining part of this study is then devoted to the development and analyze of a phase field model for the liquid-vapor flows with phase change. We propose a thermodynamic quasi-compressible formulation whose properties match the one required for the numerical study envisaged. The system of governing equations is a thermodynamically consistent regularization of the sharp interface model, that is the advantage of the di use interface models. We show that the thickness of the interface transition layer can be defined independently from the thermodynamic description of the bulk phases, a property that is numerically attractive. We derive the kinetic relation that allows to analyze the consequences of the phase field formulation on the model of the dissipative mechanisms. Finally we study the numerical resolution of the model with the help of simulations of phase transition in simple configurations as well as of isothermal bubble dynamics. (author)

  17. Boiling Heat Transfer Experiments by using Transparent Heated Microtube

    Science.gov (United States)

    Huang, Shih-Che; Kawanami, Osamu; Kawakami, Kazunari; Honda, Itsuro; Kawashima, Yousuke; Ohta, Haruhiko

    For detailed study of the relationship between boiling bubble behavior and inner wall temperature during flow boiling in microtubes, a transparent heated microtube, whose inner wall was coated with a thin gold film, was employed. Boiling behavior could be observed clearly, and the inner wall temperature of the tube was measured simultaneously with direct heating of the film. Ionized water was used as a test fluid. The experimental conditions were as follows: tube diameter, 1 mm; inlet liquid subcooling, 10 K; mass velocity, 100 kg/m2s and heat flux, up to 469 kW/m2 in the open system. As a result, the frequencies of fluctuation of the inner wall temperature and flow rate were divided into four regions. In addition, the fluctuation range of flow rate increased with increasing heat flux however, this fluctuation decreased drastically for heat flux over 212 kW/m2. The fluctuation of void fraction coincided with that of inner wall temperature.

  18. Micro-channel convective boiling heat transfer with flow instabilities

    International Nuclear Information System (INIS)

    Flow boiling heat transfer in micro-channels has attracted much interest in the past decade, and is currently a strong candidate for high performance compact heat sinks, such as those required in electronics systems, automobile air conditioning units, micro-reactors, fuel cells, etc. Currently the literature presents numerous experimental studies on two-phase heat transfer in micro-channels, providing an extensive database that covers many different fluids and operating conditions. Among the noteworthy elements that have been reported in previous studies, is the sensitivity of micro-channel evaporators to oscillatory two-phase instabilities. These periodic fluctuations in flow and pressure drop either result from the presence of upstream compressibility, or are simply due to the interaction among parallel channels in multi-port systems. An oscillating flow presents singular characteristics that are expected to produce an effect on the local heat transfer mechanisms, and thus on the estimation of the two-phase heat transfer coefficients. The present investigation illustrates results for flow boiling of refrigerants R-134a, R-236fa, and R-245fa in a 510 μm circular micro-channel, exposed to various degrees of oscillatory compressible volume instabilities. The data describe the main features of the fluctuations in the temperatures of the heated wall and fluid, and draw attention to the differences in the measured unstable time-averaged heat transfer coefficients with respect to those for stable flow boiling. (author)

  19. Measurement of film dynamics in a boiling liquid film

    International Nuclear Information System (INIS)

    Motivated by understanding the micro-hydrodynamics of boiling heat transfer and its critical heat flux (CHF), the present study investigates the boiling phenomenon in a liquid film whose dynamic thickness is recorded by a confocal optical sensor till micrometres, while the bubble dynamics of the boiling in the film is visualized by high-speed photography (100 fps). This paper is focused on statistical analysis of the thickness signals from the scoping tests from low heat flux till high heat flux (CHF). The dynamic thickness of the liquid film appears peak values, corresponding to the liquid film movements due to nucleation of bubble(s) and its growth and collapse. The maximum thickness decreases rapidly with increasing heat flux, but after 0.625 WM/m2 it keeps almost constant. It reduces again after 1.09 WM/m2 and finally reaches 105 μm prior to the CHF which occurs at 1.563 WM/m2 for the nano heater made of titanium. (author)

  20. Spray structure as generated under homogeneous flash boiling nucleation regime

    International Nuclear Information System (INIS)

    We show the effect of the initial pressure and temperature on the spatial distribution of droplets size and their velocity profile inside a spray cloud that is generated by a flash boiling mechanism under homogeneous nucleation regime. We used TSI's Phase Doppler Particle Analyzer (PDPA) to characterize the spray. We conclude that the homogeneous nucleation process is strongly affected by the initial liquid temperature while the initial pressure has only a minor effect. The spray shape is not affected by temperature or pressure under homogeneous nucleation regime. We noted that the only visible effect is in the spray opacity. Finally, homogeneous nucleation may be easily achieved by using a simple atomizer construction, and thus is potentially suitable for fuel injection systems in combustors and engines. - Highlights: • We study the characteristics of a spray that is generated by a flash boiling process. • In this study, the flash boiling process occurs under homogeneous nucleation regime. • We used Phase Doppler Particle Analyzer (PDPA) to characterize the spray. • The SMD has been found to be strongly affected by the initial liquid temperature. • Homogeneous nucleation may be easily achieved by using a simple atomizer unit

  1. Pool boiling on rectangular fins with tunnel-pore structure

    Directory of Open Access Journals (Sweden)

    Pastuszko A.

    2013-04-01

    Full Text Available Complex experimental investigations were conducted in the area of pool boiling heat transfer on extended surfaces with internal tunnels limited by perforated foil. The experiments were carried out for water and R-123 at atmospheric pressure. The tunnel surfaces were fabricated from 0.05 – 0.1 mm thick perforated copper foil (pore diameters: 0.3, 0.4, 0.5 mm sintered with mini-fins formed by 5 and 10 mm high rectangular fins and horizontal inter-fin surface. The effect of the main fin height, pore diameters and tunnel pitch on nucleate pool boiling was examined. Substantial enhancement of heat transfer coefficient was observed for the investigated surfaces. The highest increase in the heat transfer coefficient was obtained for the 10 mm high fins – about 50kW/m2K for water and 15 kW/m2K for R-123. The investigated surfaces showed boiling heat transfer coefficients similar to those of existing tunnel-pore structures.

  2. Experimental investigation and mechanistic modelling of dilute bubbly bulk boiling

    International Nuclear Information System (INIS)

    During evaporation the geometric shape of the vapour is not described using thermodynamics. In bubbly flows the bubble shape is considered spheric with small diameters and changing into various shapes upon growth. The heat and mass transfer happens at the interfacial area. The forces acting on the bubbles depend on the bubble diameter and shape. In this work the prediction of the bubble diameter and/or bubble number density in bulk boiling was considered outside the vicinity of the heat input area. Thus the boiling effects that happened inside the nearly saturated bulk were under investigation. This situation is relevant for nuclear safety analysis concerning a stagnant coolant in the spent fuel pool. In this research project a new experimental set-up to investigate was built. The experimental set-up consists of an instrumented, partly transparent, high and slender boiling container for visual observation. The direct visual observation of the boiling phenomena is necessary for the identification of basic mechanisms, which should be incorporated in the simulation model. The boiling process has been recorded by means of video images and subsequently was evaluated by digital image processing methods, and by that data concerning the characteristics of the boiling process were generated for the model development and validation. Mechanistic modelling is based on the derivation of relevant mechanisms concluded from observation, which is in line with physical knowledge. In this context two mechanisms were identified; the growth/-shrink mechanism (GSM) of the vapour bubbles and sudden increases of the bubble number density. The GSM was implemented into the CFD-Code ANSYS-CFX using the CFX Expression Language (CEL) by calculation of the internal bubble pressure using the Young-Laplace-Equation. This way a hysteresis is realised as smaller bubbles have an increased internal pressure. The sudden increases of the bubble number density are explainable by liquid super

  3. Rheological Properties and Structural Changes in Different Sections of Boiled Abalone Meat

    Institute of Scientific and Technical Information of China (English)

    GAO Xin; TANG Zhixu; ZHANG Zhaohui; Ogawa Hiroo

    2003-01-01

    Changes in tissue structures, rheological properties of cross- and vertical section boiled abalone meat were studied in relation to boiling time. The adductor muscle of abalone Haliotis discus which was removed from the shell, was boiled for 1, 2, and 3 h, respectively. Then it was cut up and separated into cross- and vertical section meat. When observed by a light microscope and a scanning electron microscope, structural changes in the myofibrils were greatest in the cross section meat compared with the vertical section meat. When boiling time was increased from 1 h to 3 h, the instantaneous modulus E0 and rupture strength of both section meat decreased gradually with increased boiling time, and no significant differences were observed between these two section meat for the same boiling time. When boiled for 1 h, the relaxation time of cross section meat was much longer than that of vertical section meat. There were no significant changes in the relaxation time of vertical section for different boiling time, but the relaxation time of cross section meat was reduced gradually with increasing boiling time. These results confirmed that the difference in rheological properties between the cross- and vertical section meat was mainly due to the denaturation level of myofibrils when heated for 1 h, as well as due to the changes in the amount of denatured proteins, and the manner in which the inner denatured protein components weve exchanged after boiling time was increased from 1 h to 3 h.

  4. Boil-off gas vapors are recovered by reliquefaction in LNG

    Energy Technology Data Exchange (ETDEWEB)

    Levay, M.; Petit, P.; Paradowski, H.

    1986-02-24

    Although great care is taken to prevent heat leaks into cryogenic equipment in LNG terminals, boil-off vapors evolve from LNG stored at thermodynamic equilibrium. The quantities of boil-off vapors may be quite considerable. They account for about 1% of the total gas quantity received and sent out at the monitor-de-bretagne LNG terminal of Gaz de France. A novel process has significantly cut boil-off vapor handling costs. It is free of technical problems which would arise from local utilization of the gas and makes boil-off recovery possible under optimum conditions. In addition, the process shows an excellent degree of reliability. Boil-off vapors have a lower heating value than the stored LNG. However, since they mainly consist of methane, their economic usefulness makes vapor recovery necessary. This boil-off gas, with widely fluctuating quantities and qualities, cannot be readily used locally. The vapors must be sent out into the grid.

  5. Needs of nuclear data for advanced light water reactor

    International Nuclear Information System (INIS)

    Hitachi has been developing medium sized ABWRs as a power source that features flexibility to meet various market needs, such as minimizing capital risks, providing a timely return on capital investments, etc. Basic design concepts of the medium sized ABWRs are 1) using the current ABWR design which has accumulated favorable construction and operation histories as a starting point; 2) utilizing standard BWR fuels which have been fabricated by proven technology; 3) achieving a rationalized design by suitably utilizing key components developed for large sized reactors. Development of the medium sized ABWRs has proceeded in a systematic, stepwise manner. The first step was to design an output scale for the 600MWe class reactor (ABWR-600), and the next step was to develop an uprating concept to extend this output scale to the 900MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. In addition, Hitachi and MHI developed an ultra small reactor, 'Package-Reactor'. About the nuclear data, for the purpose of verification of the nuclear analysis method of BWR for mixed oxide (MOX) cores, UO2 and MOX fuel critical experiments EPICURE and MISTRAL were analyzed using nuclear design codes HINES and CERES with ENDF/B nuclear data file. The critical keffs of the absorber worth experiments, the water hole worth experiments and the 2D void worth experiments agreed with those of the reference experiments within about 0.1%Δk. The root mean square differences of radial power distributions between calculation and measurement were almost less than 2.0%. The calculated reactivity worth values of the absorbers, the water hole and the 2D void agreed with the measured values within nearly experimental uncertainties. These results indicate that the nuclear analysis method of BWR in the present paper give the same accuracy for the UO2 cores and the MOX cores. (author)

  6. Transient measurement of temperature oscillation during noisy film boiling in superfluid helium II

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Noisy film boiling, which is characterized by a loud noise andsevere mechanical vibration, is a particular phenomenon of superfluid helium II (He II). Experiments have been conducted under various thermal conditions by varying the heating time th and the heat flux q, and the temperature oscillation during noisy film boiling is measured by the superconductor temperature sensors in order to understand the physical mechanism of noisy film boiling.

  7. An analytical and experimental study of pool boiling with particular reference to additives

    International Nuclear Information System (INIS)

    An experimental investigation of nucleate boiling heat transfer and critical heat flux is presented for water and various aqueous solutions boiling from horizontal stainless steel tubes and flat strips at atmospheric pressure. An integral method solution for film boiling is given and compared with existing experimental data. Analytical solutions are also obtained for the temperature profiles with periodic internal heating of a flat plate and a cylinder. (author)

  8. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  9. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  10. A review on augmentation of heat transfer in boiling using surfactants/additives

    Science.gov (United States)

    Acharya, Anil; Pise, Ashok

    2016-09-01

    Studies of heat transfer enhancement in boiling under various conditions and configurations have given different results. Understanding the boiling behaviour from these studies, literature is reviewed in terms of surface texture, heater geometry and orientation, experimental and numerical studies in presence of surfactant/additives. After understanding different behaviour in boiling, the effect of environment friendly surfactant is studied through literature review. Benchmarking of experimental procedure is done by experimenting and comparing some surfactants studied in literature.

  11. Experimental research conception of thin liquid film boiling and evaporation

    Directory of Open Access Journals (Sweden)

    Feoktistov Dmitry V.

    2015-01-01

    Full Text Available The concept of conducting the experiments for studying thin liquid film boiling and evaporation was developed. Implementing this conception on developed experimental setup, we will obtain the data on the change of liquid film thickness in thermosiphon and temperature distribution in the liquid film, also the evaporation rate of liquid film and heat transfer coefficient change will be calculated using the measured values in the experiment. Three series of preliminary experiment were conducted. As a result, the main influencing factors and their values were defined.

  12. Pool boiling in microgravity with a single specie system

    OpenAIRE

    Sagan, Michael; Colin, Catherine; Tanguy, Sébastien

    2012-01-01

    Pool boiling experiments in microgravity on the small copper plate of 1cm² have been performed in the SOURCE 2 experiment aboard the sounding Rocket Maser 12 launched on February 13th, 2012. The SOURCE 2 experiment is a small-scale tank devoted to the study of heat and mass transfers with a liquid refrigerant HFE7000 pressurised with its vapour. SOURCE 2 (SOUnding Rocket Compere Experiment) was developed in the frame of a French German space programme COMPERE (on the behaviour of propellant i...

  13. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  14. Nucleation, solvation and boiling of helium excimer clusters

    CERN Document Server

    Luna, Luis G Mendoza; Watkins, Mark J; Bonifaci, Nelly; Aitken, Frederic; von Haeften, Klaus

    2015-01-01

    Helium excimers generated by a corona discharge were investigated in the gas and normal liquid phases of helium as a function of temperature and pressure between 3.8 and 5.0 K and 0.2 and 5.6 bar. Intense fluorescence in the visible region showed the rotationally resolved $d^3\\Sigma_u^+ \\rightarrow b^3\\Pi_g$ transition of He$_2^*$. With increasing pressure, the rotational lines merged into single features. The observed pressure dependence of linewidths, shapes and lineshifts established phases of coexistence and separation of excimer-helium mixtures, providing detailed insight into nucleation, solvation and boiling of He$_2^*$-He$_n$ clusters.

  15. MTD-MFC: unified framework for investigation of diversity of boiling heat transfer curves

    Energy Technology Data Exchange (ETDEWEB)

    Shekriladze, I.G. [Georgian Technical University, Tbilisi (Georgia)], e-mail: shekri@geo.net.ge

    2009-07-01

    A keynote paper presents just the next attempt to promote a discussion of modern state of art in the field of boiling heat transfer research. It is shown how longstanding disregard of internal contradictions of applicable approaches has resulted theoretical deadlock. Alternatively, it also is shown how resolution of these contradictions opens the ways to breakthrough in boiling heat transfer theory. Basic experimental facts, physical models and correlations are reconsidered. Principal contradictions between experimental knowledge and traditional model of 'the theatre of actors' (MTA) are discussed. Crucial role of pumping effect of growing bubble (PEGB) in boiling heat transfer and hydrodynamics is shown. Basic role of control of HTC by thermodynamic conditions on nucleation sites is demonstrated and consequent model of 'the theatre of director' (MTD) is discussed. Universal MTD-based correlation of boiling HTC of all types of liquids is considered. Unified consistent research framework for developed boiling heat transfer and diverse specific boiling heat transfer regimes is outlined through supplementing MTD by so-called multifactoring concept (MFC). The latter links transition from developed boiling mode to diverse boiling curves to a phenomenon of multiplication of factors influencing HTC. The ways of further research of the boiling problem are discussed. (author)

  16. Experimental Evidence of the Vapor Recoil Mechanism in the Boiling Crisis

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Beysens, D

    2016-01-01

    Boiling crisis experiments are carried out in the vicinity of the liquid-gas critical point of H2. A magnetic gravity compensation setup is used to enable nucleate boiling at near critical pressure. The measurements of the critical heat flux that defines the threshold for the boiling crisis are carried out as a function of the distance from the critical point. The obtained power law behavior and the boiling crisis dynamics agree with the predictions of the vapor recoil mechanism and disagree with the classical vapor column mechanism.

  17. Experimental study on the pool boiling CHF enhancement using magnetic fluid

    International Nuclear Information System (INIS)

    This paper will describe the effects of magnetic fluid on CHF enhancement of pool boiling. In order to evaluate the effects as nanoparticle characteristic of magnetic fluid, we compared the CHF values of pool boiling experiment between magnetic fluid and other nanofluids with several volume concentrations. SEM(Scanning Electron Microscope) images were obtained to explain CHF enhancement through the effect of the deposited nanoparticles, which can change the surface wettability, during the pool boiling experiment. Lastly, the analysis for bubble formation in pool boiling using image processing was performed to demonstrate between the characteristics of bubble formation and CHF enhancement. (author)

  18. Two-dimensional simulation of the downcomer boiling experiment using the CUPID code

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the DOBO (Downcomer Boiling) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change and a circulation of liquid accelerated by bubbles. The two-phase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling. (author)

  19. Burnout in a high heat-flux boiling system with an impinging jet

    International Nuclear Information System (INIS)

    An experimental study has been made on the fully-developed nucleate boiling at atmospheric pressure in a simple forced-convection boiling system, which consists of a heated flat surface and a small, high-speed jet of water or of freon-113 impinging on the heated surface. A generalized correlation for burnout heat flux data, that is applied to either water or freon-113 is successfully evolved, and it is shown that surface tension has an important role for the onset of burnout phenomenon, not only in the ordinary pool boiling, but also in the present boiling system with a forced flow. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Baglietto, E.; Pointer, W. D.

    2016-08-01

    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 todays light water reactors and their fuel. An important part of the CASL mission is the development of a next generation thermal hydraulics simulation capability, integrating the history of engineering models based on based on experimental experience with the computing technology of the future. (Author)

  1. Development and Capabilities of ISS Flow Boiling and Condensation Experiment

    Science.gov (United States)

    Nahra, Henry; Hasan, Mohammad; Balasubramaniam, R.; Patania, Michelle; Hall, Nancy; Wagner, James; Mackey, Jeffrey; Frankenfield, Bruce; Hauser, Daniel; Harpster, George; Nawrocki, David; Clapper, Randy; Kolacz, John; Butcher, Robert; May, Rochelle; Chao, David; Mudawar, Issam; Kharangate, Chirag R.; O'Neill, Lucas E.

    2015-01-01

    An experimental facility to perform flow boiling and condensation experiments in long duration microgravity environment is being designed for operation on the International Space Station (ISS). This work describes the design of the subsystems of the FBCE including the Fluid subsystem modules, data acquisition, controls, and diagnostics. Subsystems and components are designed within the constraints of the ISS Fluid Integrated Rack in terms of power availability, cooling capability, mass and volume, and most importantly the safety requirements. In this work we present the results of ground-based performance testing of the FBCE subsystem modules and test module which consist of the two condensation modules and the flow boiling module. During this testing, we evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, heat loss from different modules and components, and performance of the test modules. These results will be used in the refinement of the flight system design and build-up of the FBCE which is manifested for flight in late 2017-early 2018.

  2. Effect of size sprinkled heat exchange surface on developing boiling

    Directory of Open Access Journals (Sweden)

    Petr Kracík

    2016-06-01

    Full Text Available This article presents research of sprinkled heat exchangers. This type of research has become rather topical in relation to sea water desalination. This process uses sprinkling of exchangers which rapidly separates vapour phase from a liquid phase. Applications help better utilize low-potential heat which is commonly wasted in utility systems. Low-potential heat may increase utilization of primary materials. Our ambition is to analyse and describe the whole sprinkled exchanger. Two heat exchangers were tested with a similar tube pitch: heat exchanger no. 1 had a four-tube bundle and heat exchanger no. 2 had eight-tube bundle. Efforts were made to maintain similar physical characteristics. They were tested at two flow rates (ca 0.07 and 0.11 kg s−1 m−1 and progress of boiling on the bundle was observed. Initial pressure was ca 10 kPa (abs at which no liquid was boiling at any part of the exchanger; the pressure was then lowered. Other input parameters were roughly similar for both flow rates. Temperature of heating water was ca 50°C at a constant flow rate of ca 7.2 L min−1. Results of our experiments provide optimum parameters for the given conditions for both tube bundles.

  3. Subcooled pool boiling on thin wire in microgravity

    Science.gov (United States)

    Zhao, J. F.; Wan, S. X.; Liu, G.; Yan, N.; Hu, W. R.

    2009-01-01

    A new set of experimental data of subcooled pool boiling on a thin wire in microgravity aboard the 22nd Chinese recoverable satellite is reported in the present paper. The temperature-controlled heating method is used. The results of the experiments in normal gravity before and after the flight experiment are also presented, and compared with those in microgravity. The working fluid is degassed R113 at 0.1 MPa and subcooled by 26C nominally. A thin platinum wire of 60μm in diameter and 30 mm in length is simultaneously used as heater and thermometer. It is found that the heat transfer of nucleate pool boiling is slightly enhanced in microgravity comparing with those in normal gravity. It is also found that the correlation of Lienhard and Dhir can predict the CHF with good agreement, although the range of the dimensionless radius is extended by three or more decades above the originally set limit. Three critical bubble diameters are observed in microgravity, which divide the observed vapor bubbles into four regimes with different sizes. Considering the Marangoni effect, a qualitative model is proposed to reveal the mechanism underlying the bubble departure processes, and a quantitative agreement can also be acquired.

  4. Modelling of boiling bubbly flows using a polydisperse approach

    International Nuclear Information System (INIS)

    The objective of this work was to improve the modelling of boiling bubbly flows.We focused on the modelling of the polydisperse aspect of a bubble population, i.e. the fact that bubbles have different sizes and different velocities. The multi-size aspect of a bubble population can originate from various mechanisms. For the bubbly flows we are interested in, bubble coalescence, bubble break-up, phase change kinematics and/or gas compressibility inside the bubbles can be mentioned. Since, bubble velocity depends on bubble size, the bubble size spectrum also leads to a bubble velocity spectrum. An averaged model especially dedicated to dispersed flows is introduced in this thesis. Closure of averaged interphase transfer terms are written in a polydisperse framework, i.e. using a distribution function of the bubble sizes and velocities. A quadratic law and a cubic law are here proposed for the modelling of the size distribution function, whose evolution in space and time is then obtained with the use of the moment method. Our averaged model has been implemented in the NEPTUNE-CFD computation code in order to simulate the DEBORA experiment. The ability of our model to deal with sub-cooled boiling flows has therefore been evaluated. (author)

  5. High Heat Flux Burnout in Subcooled Flow Boiling

    Institute of Scientific and Technical Information of China (English)

    G.P.Celata; M.Cumo; 等

    1995-01-01

    The paper reports the results of an experimental research carried out at the Heat transfer divison 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 2.5to 8.0 mm),heated length(10 and 15cm) ,liquid velocity (from 2 to 40m/s),exit pressure(from atmospheric to 5.0 MPa),inlet temperature(from 30 to 80℃),channel orientation (vertical and horizontal),A maximum CHF value of 60.6MW/m2 has been obtained under the following conditions:Tin-30°,p=2.5MPa,u=40m/s,D=2.5mm(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.

  6. Critical Heat Flux during Flow Boiling Experiment with Surfactant Solutions

    International Nuclear Information System (INIS)

    Some additives enhance heat transfer, although, the magnitude and mechanism of enhancement are not consistent or clearly understood. A low concentration of surfactant can also reduce the solution's surface tension considerably, and its level of reduction depends on the amount and type of surfactant present in solution. The surfactant concentrations are usually low enough that the addition of surfactant to water causes no significant change in saturation temperature and most other physical properties, except viscosity and surface tension. Reduced surface tension influences the activation of nucleation sites, bubble growth and dynamics, affecting the boiling heat transfer coefficient. Surfactants effect on CHF (Critical Heat Flux) was determined during flow boiling at atmospheric pressure in closed loop filled with water solutions of tri-sodium phosphate (TSP, Na3PO4.12H2O). TSP was added to the containment sump water to adjust pH level during accidents in nuclear power plants. CHF was measured for four water surfactant solutions at different mass fluxes (100 - 500 kg/m2sec) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Wettability was determined by measuring the contact angle at different concentration cases that will substantiate any CHF increase

  7. On-line monitoring of boiling crevice chemistry evolution

    International Nuclear Information System (INIS)

    In a locally restricted geometry on the secondary side of steam generator (SG) in a pressurized water reactor (PWR), impurities in bulk water can be concentrated by boiling process to extreme pH that may then accelerate the corrosion of tubing and adjacent materials. To simulate a real SG tubesheet crevice, a high temperature/high pressure (HT/HP) crevice simulation system was constructed. Primary water was pumped at a high flow rate through a 3/4'' outer-diameter tubing and a crevice section was made on the outer diameter (OD) side of the tubing. The simulated crevice area was monitored with thermocouples and electrodes for the measurement of temperature and electrochemical corrosion potential (ECP), respectively, in the crevice as well as free span. A secondary solution composed of 50 ppm Na and 200 ppb hydrogen (H2) was supplied at a flow rate of about 4 L/hr. In an open tubesheet crevice with 0.15 mm radial gap and 40 mm depth, axial distributions of temperature and ECP were measured as a function of time and available superheat. Sodium hydroxide (NaOH) concentration process in the crevice and the resultant evolution of crevice boiling regions were characterized from temperature and ECP data. Measured data for an open crevice showed a similar behavior to predictions by a thermodynamic equilibrium code. Magnetite-packed crevice had much longer time to reach a steady state than open crevice. (authors)

  8. Radiolysis effects in sub-cooled nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Dickinson, S.; Henshaw, J.; Tuson, A.; Sims, H.E. [AEA Technology (United Kingdom)

    2002-07-01

    A hydrogen depleted region may form in the water during bubble formation when boiling occurs in a PWR. This would arise from stripping of gases into the steam phase. The depleted water may then become oxidising due to radiolysis forming H{sub 2}O{sub 2}. The presence of radiolytic oxidising conditions is one of the mechanisms proposed to explain deposits formed in Axial Offset Anomalies. This work describes a model that has been developed to examine this behaviour. The model deals with bubble growth and material transport as well as the radiolysis chemistry. The model simulates diffusion of species through the gas/liquid boundary layer. The appropriate mass conservation equations for this problem are described and the results of their numerical solution discussed. This model indicates the importance of the assumed boundary conditions on the results of the calculations. These boundary conditions are discussed in detail and the most appropriate ones for the actual reactor situation are outlined. The conclusion of this modelling study is that at normal PWR operating conditions of 40 cc H{sub 2} (STP) kg{sup -1} it is unlikely that radiolysis in a subcooled boiling region would be important. The situation is more ambiguous at the 1 to 5 cc H{sub 2} (STP) kg{sup -1} range. (author)

  9. Nucleate boiling pressure drop in an annulus: Book 5

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux. This document consists solely of the plato file index from 11/87 to 11/90

  10. Effect of superheat and electric field on saturated film boiling

    Science.gov (United States)

    Pandey, Vinod; Biswas, Gautam; Dalal, Amaresh

    2016-05-01

    The objective of this investigation is to study the influence of superheat temperature and applied uniform electric field across the liquid-vapor interface during film boiling using a coupled level set and volume of fluid algorithm. The hydrodynamics of bubble growth, detachment, and its morphological variation with electrohydrodynamic forces are studied considering the medium to be incompressible, viscous, and perfectly dielectric at near critical pressure. The transition in interfacial instability behavior occurs with increase in superheat, the bubble release being periodic both in space and time. Discrete bubble growth occurs at a smaller superheat whereas vapor columns form at the higher superheat values. Destabilization of interfacial motion due to applied electric field results in decrease in bubble separation distance and increase in bubble release rate culminating in enhanced heat transfer rate. A comparison of maximum bubble height owing to application of different intensities of electric field is performed at a smaller superheat. The change in dynamics of bubble growth due to increasing superheat at a high intensity of electric field is studied. The effect of increasing intensity of electric field on the heat transfer rate at different superheats is determined. The boiling characteristic is found to be influenced significantly only above a minimum critical intensity of the electric field.

  11. Experimental Research on Flash Boiling Spray of Dimethyl Ether

    Institute of Scientific and Technical Information of China (English)

    Peng Zhang

    2014-01-01

    The high-speed digital imaging technique is applied to observe the developing process of flash boiling spray of dimethyl ether at low ambient pressure, and the effects of nozzle opening pressure and nozzle hole diameter on the spray shape, spray tip penetration and spray angle during the injection are investigated. The experimental results show that the time when the vortex ring structure of flash boiling spray forms and its developing process are determined by the combined action of the bubble growth and breakup in the spray and the air drag on the leading end of spray;with the enhancement of nozzle opening pressure, the spray tip penetration increases and the spray angle decreases. The influence of nozzle hole diameter on the spray tip penetration is relatively complicated, the spray tip penetration is longer with a smaller nozzle hole diameter at the early stage of injection, while the situation is just opposite at the later stage of injection. This paper establishes that the variation of spray angle is consistent with that of nozzle hole diameter.

  12. Drag reduction of flow boiling with polymer additives

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The drag-reducing effect of polymer additive aqueous solution was investigated in flow boiling, and the polymer additives were two kinds of polyacrylamide (PAM) with relative molecular mass about 2.56×106 and 8.55×106. The frictional pressure drop was calculated according to the measured total pressure drop. The results show that the flow drag of flow boiling is reduced by adding a small amount of PAM to water when heat flux is in the range of 15.1 kW*m-2 to 47.0 kW*m-2, when the mass fraction of PAM is higher than 2.0×10-5, the drag-reducing effect is obvious. Drag-reducing effect of PAM, whose relative molecular mass is 8.55×106, is slightly better than that of 2.56×106 at the same mass fraction, and the greater the flow rate of the additive solution, the better the effect of the drag reduction.

  13. Nucleate Pool Boiling of Pure Liquids and Binary Mixtures :Part I—Analytical Model for Boiling Heat Transfer of Pure Liquids on Smooth Tubes

    Institute of Scientific and Technical Information of China (English)

    GuoqingWang; YingkeTan; 等

    1996-01-01

    A mechanism is proposed for nucleate pool boiling heat transfer along with a general model for both pure liquids and binary mixtrues.A combined physical model of bubble growth is also proposed along with a corresponding bubble growth model for pure liquids on smooth tubes.Using the general model and the bubble growth model for pure liquids,an analyticasl model for nucleate pool boiling heat transfer of pure liquids on smooth tubes is developed.

  14. 46 CFR 154.709 - Cargo boil-off as fuel: Gas detection equipment.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Gas detection equipment. 154.709 Section 154.709 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS... Equipment Cargo Pressure and Temperature Control § 154.709 Cargo boil-off as fuel: Gas detection...

  15. 46 CFR 154.708 - Cargo boil-off as fuel: Valves.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Valves. 154.708 Section 154.708 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY... Pressure and Temperature Control § 154.708 Cargo boil-off as fuel: Valves. (a) Gas fuel lines to the...

  16. 46 CFR 154.707 - Cargo boil-off as fuel: Ventilation.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Ventilation. 154.707 Section 154.707 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES... Equipment Cargo Pressure and Temperature Control § 154.707 Cargo boil-off as fuel: Ventilation. (a)...

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

    Science.gov (United States)

    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.

  18. Thermodynamic and Gasdynamic Aspects of a Boiling Liquid Expanding Vapour Explosion

    NARCIS (Netherlands)

    Xie, M.

    2013-01-01

    The risk of explosion due to rupture of a tank filled with pressurized liquefied gas (PLG) is one of the risks to be considered in the context of studies on tunnel safety. When a vessel containing liquid well above its boiling point at normal atmospheric pressure fails catastrophically a Boiling Liq

  19. The Gibbs Energy Basis and Construction of Boiling Point Diagrams in Binary Systems

    Science.gov (United States)

    Smith, Norman O.

    2004-01-01

    An illustration of how excess Gibbs energies of the components in binary systems can be used to construct boiling point diagrams is given. The underlying causes of the various types of behavior of the systems in terms of intermolecular forces and the method of calculating the coexisting liquid and vapor compositions in boiling point diagrams with…

  20. Teaching Structure-Property Relationships: Investigating Molecular Structure and Boiling Point

    Science.gov (United States)

    Murphy, Peter M.

    2007-01-01

    A concise, well-organized table of the boiling points of 392 organic compounds has facilitated inquiry-based instruction in multiple scientific principles. Many individual or group learning activities can be derived from the tabulated data of molecular structure and boiling point based on the instructor's education objectives and the students'…

  1. 46 CFR 154.706 - Cargo boil-off as fuel: Fuel lines.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Fuel lines. 154.706 Section 154... Equipment Cargo Pressure and Temperature Control § 154.706 Cargo boil-off as fuel: Fuel lines. (a) Gas fuel lines must not pass through accommodation, service, or control spaces. Each gas fuel line...

  2. The cholesterol-raising factor from boiled coffee does not pass a paper filter.

    NARCIS (Netherlands)

    Dusseldorp, van M.; Katan, M.B.; Vliet, van T.; Demacker, P.N.M.; Stalenhoef, A.F.H.

    1991-01-01

    Previous studies have indicated that consumption of boiled coffee raises total and low density lipoprotein (LDL) cholesterol, whereas drip-filtered coffee does not. We have tested the effect on serum lipids of consumed coffee that was first boiled and then filtered through commercial paper coffee fi

  3. Microwave super-heated boiling of organic liquids: Origin, effect and application

    NARCIS (Netherlands)

    Chemat, F.; Esveld, E.

    2001-01-01

    This paper reports the state of the art of the microwave super-heated boiling phenomenon. When a liquid is heated by microwaves, the temperature increases rapidly to reach a steady temperature while refluxing. It happens that this steady state temperature can be up to 40 K higher than the boiling po

  4. Theoretical modeling of CHF for near-saturated pool boiling and flow boiling from short heaters using the interfacial lift-off criterion

    Energy Technology Data Exchange (ETDEWEB)

    Mudawar, I.; Galloway, J.E.; Gersey, C.O. [Purdue Univ., West Lafayette, IN (United States)] [and others

    1995-12-31

    Pool boiling and flow boiling were examined for near-saturated bulk conditions in order to determine the critical heat flux (CHF) trigger mechanism for each. Photographic studies of the wall region revealed features common to both situations. At fluxes below CHF, the vapor coalesces into a wavy layer which permits wetting only in wetting fronts, the portions of the liquid-vapor interface which contact the wall as a result of the interfacial waviness. Close examination of the interfacial features revealed the waves are generated from the lower edge of the heater in pool boiling and the heater`s upstream region in flow boiling. Wavelengths follow predictions based upon the Kelvin-Helmholtz instability criterion. Critical heat flux in both cases occurs when the pressure force exerted upon the interface due to interfacial curvature, which tends to preserve interfacial contact with the wall prior to CHF, is overcome by the momentum of vapor at the site of the first wetting front, causing the interface to lift away from the wall. It is shown this interfacial lift-off criterion facilitates accurate theoretical modeling of CHF in pool boiling and in flow boiling in both straight and curved channels.

  5. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  6. Fouling of Structured Surfaces during Pool Boiling of Aqueous Solutions

    International Nuclear Information System (INIS)

    Bubble characteristics in terms of density, size, frequency and motion are key factors that contribute to the superiority of nucleate pool boiling over the other modes of heat transfer. Nevertheless, if heat transfer occurs in an environment which is prone to fouling, the very same parameters may lead to accelerated deposit formation due to concentration effects beneath the growing bubbles. This has led heat exchanger designers frequently to maintain the surface temperature below the boiling point if fouling occurs, e.g. in thermal seawater desalination plants. The present study investigates the crystallization fouling of various structured surfaces during nucleate pool boiling of CaSO4 solutions to shed light into their fouling behaviour compared with that of plain surfaces for the same operating conditions. As for the experimental part, a comprehensive set of clean and fouling experiments was performed rigorously. The structured tubes included low finned tubes of different fin densities, heights and materials and re-entrant cavity Turbo-B tube types.The fouling experiments were carried out at atmospheric pressure for different heat fluxes ranging from 100 to 300 k W/m2 and CaSO4 concentrations of 1.2 and 1.6 g/L. For the sake of comparison, similar runs were performed on plain stainless steel and copper tubes.Overall for the finned tubes, the experimental results showed a significant reduction of fouling resistances of up to 95% compared to those of the stainless steel and copper plain tubes. In addition, the scale formation that occurred on finned tubes was primarily a scattered and thin crystalline layer which differs significantly from those of plain tubes which suffered from a thick and homogenous layer of deposit with strong adhesion. Higher fin densities and lower fin heights always led to better antifouling performance for all investigated finned tubes. It was also shown that the surface material strongly affects the scale formation of finned tubes i

  7. High conversion pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–233U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–233U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm3, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore, some means of

  8. Contribution to the development of a Local Predictive Approach of the boiling crisis

    International Nuclear Information System (INIS)

    EDF aims at developing a 'Local Predictive Approach' of the boiling crisis for PWR core configurations, i.e. an approach resulting in (empirical) critical heat flux predictors based on local parameters provided by NEPTUNE-CFD code (for boiling bubbly flows, only in a first stage). Within this general framework, this PhD work consisted in assess one modelling of NEPTUNE-CFD code selected to simulate boiling bubble flows, then improve it. The latter objective led us to focus on the mechanistic modelling of subcooled nucleate boiling in forced convection. After a literature review, we identified physical improvements to be accounted for, especially with respect to bubble sliding phenomenon along the heated wall. Subsequently, we developed a force balance model in order to provide needed closure laws related to bubble detachment diameter from the nucleation site and lift-off bubble diameter from the wall. A new boiling model including such developments was eventually proposed, and preliminary assessed. (author)

  9. Evaluation of correlations of flow boiling heat transfer of R22 in horizontal channels.

    Science.gov (United States)

    Zhou, Zhanru; Fang, Xiande; Li, Dingkun

    2013-01-01

    The calculation of two-phase flow boiling heat transfer of R22 in channels is required in a variety of applications, such as chemical process cooling systems, refrigeration, and air conditioning. A number of correlations for flow boiling heat transfer in channels have been proposed. This work evaluates the existing correlations for flow boiling heat transfer coefficient with 1669 experimental data points of flow boiling heat transfer of R22 collected from 18 published papers. The top two correlations for R22 are those of Liu and Winterton (1991) and Fang (2013), with the mean absolute deviation of 32.7% and 32.8%, respectively. More studies should be carried out to develop better ones. Effects of channel dimension and vapor quality on heat transfer are analyzed, and the results provide valuable information for further research in the correlation of two-phase flow boiling heat transfer of R22 in channels.

  10. A fractal study for nucleate pool boiling heat transfer of nanofluids

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    In this paper, a fractal model for nucleate pool boiling heat transfer of nanofluids is developed based on the fractal distribution of nanoparticles and nucleation sites on boiling surfaces. The model shows the dependences of the heat flux on nanoparticle size and the nanoparticle volume fraction of the suspension, the fractal dimension of the nanoparticle and nucleation site, temperature of nanofluids and properties of fluids. The fractal model predictions show that the natural convection stage continues relatively longer in the case of nanofluids. The addition of nanoparticles causes a decrease of the pool nucleate boiling heat transfer. The nucleate pool boiling heat transfer coefficient is decreased by increasing particle concentration. An excellent agreement between the proposed model predictions and experimental data is found. The validity of the fractal model for nucleate pool boiling heat transfer is thus verified.

  11. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

    An evaluation of the off-gas systems for all 25 operating Boiling Water Reactors (BWR) was made to determine the adequacy of their design and operating procedures to reduce the probability of off-gas detonations. The results of the evaluations are that, of the 25 operable units, 13 meet all the acceptance criteria. The other 12 units do not have the features needed to meet the criteria, but have been judged to have, or are committed to provide, features which give reasonable assurance that the potential for external off-gas detonations is minimized. The 12 units which did not originally meet the criteria are aware of the potential hazards associated with off-gas detonations and have agreed to take action to minimize the probability of future detonations

  12. Boiling visualization on vertical fins with tunnel-pore structures

    Directory of Open Access Journals (Sweden)

    Kaniowski Robert

    2012-04-01

    Full Text Available The paper presents experimental studies of nucleate boiling heat transfer from a system of connected horizontal and vertical subsurface tunnels. The experiments were carried out for water at atmospheric pressure. The tunnel external covers were manufactured out of perforated copper foil (holes diameter 0.3 mm, sintered with the mini-fins, formed on the vertical side of the 10 mm high rectangular fins and horizontal inter-fin surface. The image acquisition speed was 493 fps (at resolution 400 × 300 pixels with Photonfocus PHOT MV-D1024-160-CL camera. Visualization investigations aimed to identify nucleation sites and flow patterns and to determine the bubble departure diameter and frequency at various superheats for vertical tunnels. At low superheat vapor bubbles are generated nearly exclusively by the vertical tunnel. At medium values of superheat, pores of the horizontal tunnel activate.

  13. Channel-type nuclear reactor with a boiling coolant

    International Nuclear Information System (INIS)

    The invention is aimed at increasing the channel-type reactor safety, in particular, RBMK-type reactors, during accidents resulting in the coolant circulation discontinuation. The reactor core is assembled of vertial technological channels connected in parallel between distributing group collectors and drum-separator. Each technological channel contains a high pressure tube, a fuel assembly with fuel elements and a storage vessel located above the fuel assembly which is filled with water at saturation temperature in the normal operation regime. After dehydration of channels in the course of accident the boiling water from storage vessel is ejected into them. So the device described allows one to reduce the fuel element can temperature in the course of accidents connected with the coolant circulation discontinuation and so to increase the plant safety level

  14. Flow boiling of water on nanocoated surfaces in a microchannel

    CERN Document Server

    Phan, Hai Trieu; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2010-01-01

    Experiments were performed to study the effects of surface wettability on flow boiling of water at atmospheric pressure. The test channel is a single rectangular channel 0.5 mm high, 5 mm wide and 180 mm long. The mass flux was set at 100 kg/m2 s and the base heat flux varied from 30 to 80 kW/m2. Water enters the test channel under subcooled conditions. The samples are silicone oxide (SiOx), titanium (Ti), diamond-like carbon (DLC) and carbon-doped silicon oxide (SiOC) surfaces with static contact angles of 26{\\deg}, 49{\\deg}, 63{\\deg} and 103{\\deg}, respectively. The results show significant impacts of surface wettability on heat transfer coefficient.

  15. Interface oscillation of subcooled flow boiling in locally heated microchannels

    Science.gov (United States)

    Liu, J. T.; Peng, X. F.

    2009-02-01

    An investigation was conducted to understand flow boiling of subcooled de-ionized water in locally heated parallel microchannels. High-speed visualization technology was employed to visually observe the transient phase change process in an individual microchannel. Signal analysis method was employed in studying the interface movement and phase change process. The phase change at locally heated condition was different from those at entirely heated condition where elongated bubble(s) stayed quasi-stable for a long time without venting out. Diversified and intensive interface oscillation was observed occurring on both of the upstream and downstream bubble caps. Evaporation and condensation modes were characterized with distinguished oscillation frequencies. The film-driven oscillations of both evaporating and condensing interfaces generally operated at higher frequencies than the oscillations driven by nucleation or dropwise condensation.

  16. Pool boiling of nanoparticle-modified surface with interlaced wettability

    KAUST Repository

    Hsu, Chin-Chi

    2012-01-01

    This study investigated the pool boiling heat transfer under heating surfaces with various interlaced wettability. Nano-silica particles were used as the coating element to vary the interlaced wettability of the surface. The experimental results revealed that when the wettability of a surface is uniform, the critical heat flux increases with the more wettable surface; however, when the wettability of a surface is modified interlacedly, regardless of whether the modified region becomes more hydrophilic or hydrophobic, the critical heat flux is consistently higher than that of the isotropic surface. In addition, this study observed that critical heat flux was higher when the contact angle difference between the plain surface and the modified region was smaller. © 2012 Hsu et al.

  17. Flow Structures Around Micro-bubbles During Subcooled Nucleate Boiling

    Institute of Scientific and Technical Information of China (English)

    WANG Hao; PENG Xiao-Feng; David M. Christopher; WANG Bu-Xuan

    2005-01-01

    The flow structures were investigated around micro bubbles on extremely thin wires during subcooled nucleate boiling. Jet flows emanating from the bubbles were observed visually with the fluid field measurement using high-speed photography and a PIV system. The jet flows induced a strong pumping effect around a bubble. The multi-jet structure was further observed experimentally, indicating the evolution of flow structure around micro bubbles. Numerical simulations explore that the jet flows were induced by a strong Marangoni effect due to high temperature gradients near the wire. The bubble interface with multi-jet structure has abnormal temperature distribution such that the coolest parts were observed at two sides of a bubble extending into the subcooled bulk liquid rather than at the top. Evaporation and condensation on the bubble interface play important roles not only in controlling the intensity of the jet flow, but also in bringing out the multi-jet structure.

  18. Dynamic simulation of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    For the application of modern control theory, specifically optimal control, to the boiling water reactor, it is necessary to have a linear model that is validated. The nonlinear model of the BWR derived on the basis of physical laws and empirical relations is linearized around an operating point and the model if verified against experimental results by simulating various tests such as the pressure transient test, change in power to recirculating pump etc. The transport delay occurring in the model is approximated by various representations and the results are compared with the exact delay representation. Validation such as discussed in the paper forms the basis for devising appropriate control strategies in the presence of disturbances. (author)

  19. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  20. Theoretical prediction method of subcooled flow boiling CHF

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Chang, Soon Heung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

  1. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  2. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  3. Nucleate boiling pressure drop in an annulus: Book 3

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements

  4. Nucleate boiling pressure drop in an annulus: Book 6

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of a summary of temperature measurements to include recorded minima, maxima, averages and standard deviations.

  5. Nucleate boiling pressure drop in an annulus: Book 7

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists solely of tables of temperature measurements; minima, maxima, averages and standard deviations being measured.

  6. Nucleate boiling pressure drop in an annulus: Book 4

    Energy Technology Data Exchange (ETDEWEB)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  7. Nucleate boiling pressure drop in an annulus: Book 8

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of tables of temperature measurements.

  8. Nucleate boiling pressure drop in an annulus: Book 3

    Energy Technology Data Exchange (ETDEWEB)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  9. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Boiling water nuclear reactors may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate due to the tight coupling of flow to power, especially under gravity-driven circulation. In order to predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model has been developed for a typical boiling water reactor. Using this tool it has been demonstrated that density waxes may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases have been analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. From that study it has been concluded that two-phase friction controls the extent of the oscillation and that the existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from case to case. It has also been determined that higher dimensional nuclear feedback models reduce the extent of the oscillation. It has also been confirmed from a nonlinear dynamic standpoint that the birth of this oscillation may be described as a Hopf Bifurcation

  10. Nucleate boiling pressure drop in an annulus: Book 2

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux

  11. Technical and QA plan: Boiling behavior during flow instability

    International Nuclear Information System (INIS)

    The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may occur. This localized boiling can restrict coolant flow and the ability to transfer heat out of the reactor system. The maximum operating power for the reactor may be limited by how the coolant system reacts to a flow instability. One of the methods to assure safe operation during a reducing flow transient, is to operate at a power level below that necessary to initiate a flow excursion. Several correlations have been used to predict the conditions which will proceed a flow excursion. These correlations rely on the steady state behavior of the coolant and are based on steady-state testing. There are two significant points which this project will try to identify. The first is when vapor first forms on the channel surface. This might be designated as the Nucleate Vapor Transition. (Steady state equivalent is ONB). The second is when the vapor formation rate is large enough to lead to flow instability and thermal excursion. This point might be designated as the Significant Vapor Transition. (Steady state equivalent is OSV). A correlation will be developed to relate established steady state relations with the behavior of transient systems

  12. Nucleate boiling pressure drop in an annulus: Book 4

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements

  13. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.''...

  14. On-line test of power distribution prediction system for boiling water reactors

    International Nuclear Information System (INIS)

    A power distribution prediction system for boiling water reactors has been developed and its on-line performance test has proceeded at an operating commercial reactor. This system predicts the power distribution or thermal margin in advance of control rod operations and core flow rate change. This system consists of an on-line computer system, an operator's console with a color cathode-ray tube, and plant data input devices. The main functions of this system are present power distribution monitoring, power distribution prediction, and power-up trajectory prediction. The calculation method is based on a simplified nuclear thermal-hydraulic calculation, which is combined with a method of model identification to the actual reactor core state. It has been ascertained by the on-line test that the predicted power distribution (readings of traversing in-core probe) agrees with the measured data within 6% root-mean-square. The computing time required for one prediction calculation step is less than or equal to 1.5 min by an HIDIC-80 on-line computer

  15. Boiling Water Reactor Loading Pattern Optimization Using Simple Linear Perturbation and Modified Tabu Search Methods

    International Nuclear Information System (INIS)

    An automated system for designing a loading pattern (LP) for boiling water reactors (BWRs) given a reference LP and control rod (CR) sequence has been developed. This system employs the advanced nodal code SIMULATE-3 and a BWR LP optimization code FINELOAD-3, which uses a simple linear perturbation method and a modified Tabu search method to select potential optimized LP candidates. Both of these unique methods of FINELOAD-3 were developed to achieve an effective BWR LP optimization strategy and to have high computational efficiency. FINELOAD-3 also adjusts deep CR positions to compensate for the core reactivity deviation caused by fuel shuffling. The objective function is to maximize the end-of-cycle core reactivity while satisfying the specified thermal margins and cold shutdown margin constraints. This optimization system realized the practical application for real BWR LP design. Computer time needed to obtain an optimized LP for a typical BWR/5 octant core with 15 depletion steps is ∼4 h using an engineering workstation. This system was extensively tested for real BWR reload core designs and showed that the developed LPs using this system are equivalent or better than the manually optimized LPs

  16. Gold Deposition by Boiling or Cooling Without Boiling: Genesis of the Sangchon Gold Deposits, Hadong Area, South Korea

    Institute of Scientific and Technical Information of China (English)

    Maeng - Eon PARK; Kyu - Youl SUNG; Seong - Taek YUN

    2001-01-01

    In order to understand the mechanism(s) of gold precipitation in the anorthosite- hosted Sangchon gold deposits in the Hadong area, Korea, chemical speciation and reaction path calculations were accomplished by geochemical modeling.The modeling consisted of three- step procedures: reaction with anorthosite, then the simple cooling of the reacted fluid,and finally the boiling of metalliferous fluid. The principal vein minerals of the Sangchon deposits consist of quartz, sericite,kaolinite, pyrite, galena, chalcopyrite, sphalerite and acanthite. The sulfide mineralization is typically zoned from pyrite (preferentially at vein margins) to galena and sphalerite (toward vein center). Electrum is intimately associated with pyrite + chalcopyrite and sphalerite. By comparing the results of modeling with the observed mineral assemblages and paragenesis,the most appropriate evolution path of ore fluids was suggested as follow: reaction of a single fluid with anorthosite at 300℃,then the isobaric cooling of the fluid at temperatures from 2500° to 100℃3 , and then the boiling and cooling of the fluid due to the decrease of pressure and temperature. Calculations also show that all of the observed alteration minerals formed due to fluid - anorthosite interaction at early period, whereas most of sulfides and electrum were precipitated mainly due to cooling.The abundance of gold in veins depends critically on the ratio of total base metals plus iron to sulfide in the aqueous phase,because gold is transported as Au(HS)2- whose solubility is very sensitive to the sulfide activity. Our results of geochemical modeling generally fit to the observed mineral assemblages and mineral composition, indicating the usefulness of numerical simulation for elucidating the genesis of gold deposits.

  17. Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

    Energy Technology Data Exchange (ETDEWEB)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable

  18. Experimental investigation on partial pool boiling heat transfer in pure liquids

    Directory of Open Access Journals (Sweden)

    Fazel Seyed Ali Alavi

    2016-01-01

    Full Text Available Saturated partial pool boiling heat transfer has been experimentally investigated on a horizontal rod heater. The boiling liquids are including water and ethanol. The heating section is made by various materials including SS316, copper, aluminum and brass. Experiments have been performed at several degrees of surface roughness ranging between 30 and 360 micrometer average vertical deviation. The measurements are including boiling heat transfer coefficient, bubble departing diameter and frequency and also nucleation site density. The data have been compared to major existing correlations. It is shown that experimental data do not match with major correlations at the entire range of experiments with acceptable accuracy. In this article, the boiling heat transfer area has been divided in two complementary areas, the induced forced convection area and the boiling affected area. Based on two dimensionless groups, including Eötvös and Roshko numbers, a semi-empirical model have been proposed to predict the boiling heat transfer coefficient. It is shown that the proposed model provides improved performance in prediction of the boiling heat transfer coefficient in comparison with to existing correlations.

  19. A look-up table for fully developed film-boiling heat transfer

    International Nuclear Information System (INIS)

    An improved look-up table for film-boiling heat-transfer coefficients has been derived for steam-water flow inside vertical tubes. Compared to earlier versions of the look-up table, the following improvements were made: - The database has been expanded significantly. The present database contains 77,234 film-boiling data points obtained from 36 sources. - The upper limit of the thermodynamic quality range was increased from 1.2 to 2.0. The wider range was needed as non-equilibrium effects at low flows can extend well beyond the point where the thermodynamic quality equals unity. - The surface heat flux has been replaced by the surface temperature as an independent parameter. - The new look-up table is based only on fully developed film-boiling data. - The table entries at flow conditions for which no data are available is based on the best of five different film-boiling prediction methods. The new film-boiling look-up table predicts the database for fully developed film-boiling data with an overall rms error in heat-transfer coefficient of 10.56% and an average error of 1.71%. A comparison of the prediction accuracy of the look-up table with other leading film-boiling prediction methods shows that the look-up table results in a significant improvement in prediction accuracy

  20. Experiments on HFE-7100 pool boiling at atmospheric pressure in horizontal narrow spaces

    Energy Technology Data Exchange (ETDEWEB)

    Guglielmini, G.; Misale, M.; Priarone, A. [Universita degli Studi di Genova (Italy). DIPTEM - Sezione di Termoenergetica e Condizionamento Ambientale

    2009-07-01

    Experiments were performed to examine the pool boiling heat transfer and critical heat flux on a smooth copper circular surface, confined by a face-to-face parallel unheated surface, by changing the gap between the surfaces and the unheated surface diameter. Pool boiling data at atmospheric pressure were obtained for saturated HFE-7100. The gap values investigated, between the boiling surface and the adiabatic one, were s 0.5, 1.0, 2.0, 3.5 mm. To confine the boiling surface, two different Plexiglas plates were used: the former characterised by a diameter D = 60 mm, large as the overall test section support, the latter characterised by a diameter D = 30 mm, large to cover only copper boiling surface (d = 30 mm). For each configuration, boiling curves were obtained up to the thermal crisis. For both different types of confinement, it was observed that the boiling curves match at low wall superheat, except for s = 0.5 mm, 1 mm. However, at high wall superheat, a drastic reduction in heat transfer as well as CHF appears decreasing the channel width s; for all gap sizes, this reduction is less pronounced for the smaller confinement wall (D = 30 mm). Instead, at low wall superheat for gap of 0.5 and 1.0 mm, the heat transfer coefficient is higher for diameter disc of 60 mm. CHF data were also compared with a literature correlation (Misale and al., 2009). (author)

  1. Advanced, Long-Life Cryocooler Technology for Zero-Boil-Off Cryogen Storage Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Long-life, high-capacity cryocoolers are a critical need for future space systems utilizing stored cryogens. The cooling requirements for planetary and...

  2. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ...) inside the reactor building, and (3) well away from the power block. Locations inside the primary..., 2009 (74 FR 62829). On June 12, 2009 (74 FR 28112), the NRC amended its regulations to require.... Rather, the AIA rule's goal is to enhance the facility's inherent robustness at the design stage. The...

  3. Experimental Study on Convective Boiling Heat Transfer in Vertical Narrow Gap Annular Tube

    Institute of Scientific and Technical Information of China (English)

    Li Bin; He Anding; Wang Yueshe; Zhou Fangde

    2001-01-01

    Experiments are conducted to investigate the characteristics of single-phase forced-flow convection and boiling heat transfer of R113 flowing through annular tube with gap of 1, 1.5 and 2.5 mm, and also the visualization test are carried out to get two-phase flow regime. The data show that the Nusselt numbers for the narrow-gap are higher than those predicted by traditional large channel correlation and boiling heat transfer is enhanced. Based on the data obtained in this investigation, correlations for single-phase, forced convection and flow boiling in annular tube of different gap size has been developed.

  4. Criticality in the slowed-down boiling crisis at zero gravity

    OpenAIRE

    Charignon, Thomas; Lloveras Muntané, Pol Marcel; CHATAIN, Denis; Truskinovsky, Lev; Vives, Eduard; Beysens, Daniel; Nikolayev, Vadim

    2015-01-01

    Boiling crisis is a transition between nucleate and film boiling. It occurs at a threshold value of the heat flux from the heater called CHF (critical heat flux). Usually, boiling crisis studies are hindered by the high CHF and short transition duration (below 1 ms). Here we report on experiments in hydrogen near its liquid-vapor critical point, in which the CHF is low and the dynamics slow enough to be resolved. As under such conditions the surface tension is very small, the experiments are ...

  5. Bubble departure in pool and flow boiling systems: A review and latest developments

    International Nuclear Information System (INIS)

    Many of the vapor bubble departure diameter correlations for pool and flow boiling which have been proposed in the open literature are reviewed. In addition, the recent unified bubble detachment model for pool and flow boiling proposed by Zeng et al. (1992a, 1992b) is discussed. It is demonstrated that the unified model, which requires the vapor bubble growth rate as an input, is the only one which satisfactorily predicts vapor bubble departure diameters over the entire range of boiling conditions for which bubble detachment data exist

  6. The influence of three-dimensional capillary-porous coatings on heat transfer at liquid boiling

    Science.gov (United States)

    Surtaev, A. S.; Pavlenko, A. N.; Kalita, V. I.; Kuznetsov, D. V.; Komlev, D. I.; Radyuk, A. A.; Ivannikov, A. Yu.

    2016-04-01

    The process of heat transfer at pool boiling of liquid (Freon R21) on tubes with three-dimensional plasma-deposited capillary-porous coatings of various thicknesses has been experimentally studied. Comparative analysis of experimental data showed that the heat transfer coefficient for a heater tube with a 500-μm-thick porous coating is more than twice as large as that in liquid boiling on an otherwise similar uncoated tube. At the same time, no intensification of heat exchange in the regime of bubble boiling is observed on a tube with a 100-μm-thick porous coating.

  7. Boiled coffee does not increase serum cholesterol in gerbils and hamsters.

    OpenAIRE

    Mensink, R.P.; Zock, P. L.; Katan, M B; A. C. Beynen

    1992-01-01

    In contrast to drip filter coffee, boiled coffee increases the serum cholesterol level in man. To identify the substance(s) responsible for this effect, it is necessary to find an animal model sensitive to boiled coffee. In this study, three groups of 20 male gerbils and three groups of six male hamsters were fed a control diet or a control diet supplemented with either freeze-dried boiled coffee or freeze-dried filtered coffee. At the end of the 5-week feeding period serum cholesterol levels...

  8. Fourier and Wavelet Transform Analysis of Pressure Signals during Explosive Boiling

    Institute of Scientific and Technical Information of China (English)

    YIN Tie-Nan; HUAI Xiu-Lan

    2008-01-01

    @@ The transient pressure in a liquid-pool during explosive boiling of acetone is measured by a micro-pressure-measuring system.The Fast Fourier transform and continuous wavelet transform methods are applied to investigate the frequency characteristics.The results show that the dominant frequency of the explosive boiling is 0-2MHz,and the bubble cluster formed by numerous tiny bubbles departs twice.Analysis and discussions are also conducted to explain the bubble evolution during the explosive boiling.

  9. Pool boiling on the superhydrophilic surface with TiO2 nanotube arrays

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Surface with TiO2 nanotube arrays(TNTAs)is superhydrophilic and of great specific area.This paper investigates the pool boiling characteristics at the thermal interface with TNTAs.The results show that the TNTAs interface can enhance the pool boiling heat transfer compared to the pure Ti metal plate.The bubbles formed at the initial nucleation state are very small and released in higher frequency.The pool boiling heat transfer enhancement at the TNTAs interface may be attributed to the high density of nucleate site,high intrinsic heating area of nanotubes layer,superhydrophilicity and the vertically oriented nanotube structure.

  10. Comprehensive Evaluation and Prediction of Enhancement of Boiling Heat Transfer with Additives

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A model of evaluation and prediction of enhancement of boiling heat transfer with additives has been propoeed according to fuzzy fundamentals. Correlative appraisement of boiling heat transfer augmentation was done with the model based on 39 additives which were tested by the authors and other researchers. The results show that the evaluation of 35 additives is consistent with experiments, which means that the accuracy of the model is 89.7 percent. In addition, the prediction of the ability of boiling heat transfer enhancement with sodium oleate,polyethylene glycol and Tween-40 is also in good agreement with correspondent experiments.

  11. Pervaporation investigation of recovery of volatile compounds from brown crab boiling juice.

    Science.gov (United States)

    Martínez, Rodrigo; Sanz, M Teresa; Beltrán, Sagrario

    2014-10-01

    Pervaporation has been used to obtain aroma concentrates from brown crab boiling juice. The boiling juice and the obtained permeate have been analysed by Headspace Solid Phase Dynamic Extraction Gas Chromatography/Mass Spectrometry. The effect of feed temperature on the pervaporation performance of the membrane has been analysed. The permeate aroma profile, at 25 ℃ and 40 ℃, was different from that of the boiling juice. Enrichment factors for some of the volatile compounds were much lower than those obtained in model aqueous dilute solutions. Pervaporation performance can be significantly improved by modifying the permeant circuit to include two condensation stages. PMID:23897977

  12. Kandlikar third number map for flow boiling in micro-channels and micro-gravity

    Directory of Open Access Journals (Sweden)

    Awad M.M.

    2015-01-01

    Full Text Available As an extension of the recent work of Baldassari and Marengo (Baldassari C., Marengo M., Flow Boiling in Microchannels and Microgravity, Progress in Energy and Combustion Science 39 (2013 1, pp. 1-36, this note presents Kandlikar third number (K3 map for flow boiling in microchannels and microgravity. Using several data points available in the literature, Kandlikar third number (K3 map was plotted versus the hydraulic diameter (dh as the characteristic dimension for flow boiling in microchannels and microgravity. The ranges of the Kandlikar third number (K3, calculated using the hydraulic diameter (dh, are presented.

  13. Pressure measurements in boiling particle beds with water at 1 bar

    International Nuclear Information System (INIS)

    Pressures have been measured at the top and bottom of uniformly heated beds of uniform spherical particles with water boiling at atmospheric pressure. Particle sizes used vary from 0.22 to 5 mm diameter and bed heights from 50 to 150 mm. The pressures have been recorded at power levels up to dry-out. The results show how much liquid remains in a boiling bed at different power levels and how the liquid/vapour phase pressure losses vary. The results give a valuable insight into the working of a boiling bed. (author)

  14. Boiling Heat Transfer in an Acoustic Cavitation Field%声空化场下的沸腾传热

    Institute of Scientific and Technical Information of China (English)

    周定伟; 刘登瀛

    2002-01-01

    An experimental study has been carried out to investigate systematically the effects of acoustic cavitation parameters and fluid subcooling on boiling of acetone around a horizontal circular tube. The experimentalresults show that acoustic cavitation enhanced remarkably the boiling heat transfer and decreased the incipientboiling superheat and that cavitation bubbles effect on boiling heat transfer reduced with cavitation distance. Forboiling curves in a form of h-q″, elevated cavitation distance shift nucleate boiling curves to the right of the corresponding ordinary pool boiling curve. The associated mechanism of heat transfer enhancement is analyzed withthe consideration of cavitation bubble influence on vapor embryo.

  15. Construction of the first advanced BWRs with giant crawler crane module method

    International Nuclear Information System (INIS)

    The construction of No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., which are the first advanced BWRs of 1356 MW capacity and the largest class in the world, is advanced smoothly, aiming at the start of commercial operation in 1996 and 1997. The ABWR was developed aiming at further heightening the safety, operation performance and economic efficiency, and also for the construction, new construction methods were adopted and achieved effects. Hitachi Ltd. utilizes large crawler cranes up to the maximum limit, and expands the application of large section module method, and efforts are exerted for shortening the construction period, improving quality, and securing safety in the construction works. The planning and the investigation of the installation of large section module were carried out by utilizing CAD, exchanging the design information and expanding the cooperation in execution with architecture companies, in this way, the efficiency and the quality were improved. Hitachi Ltd. has promoted the construction of the turbine facilities of No. 6 plant and the nuclear reactor facilities of No. 7 plant. As the features of No. 6 and 7 plant construction, the reduction of the building size by installing internal pumps, the steel-lined reinforced concrete containment vessels and so on are mentioned. The construction schedule, and the methods of construction for No. 6 turbine building and No. 7 reactor building are reported. (K.I.)

  16. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  17. Development of micro machining tools for finishing weld joint

    International Nuclear Information System (INIS)

    GE, Hitachi and Toshiba are jointly constructing advanced boiling water reactor (ABWR) Units 6 and 7 at Kashiwazaki Kariwa Nuclear Power Plant Station, Tokyo Electric Power Co. The ABWR features enhanced operability and safety as a whole plant through simplicity and improved performance. To achieve these improvement, one of the key features of technical innovation adopted in the ABWR design, ten reactor internal pumps (RIP) are adopted as the reactor recirculation system. The RIP casing to hold the RIP constituting the primary pressure boundary together with a RPV is welded to the nozzle on a RPV lower shell with Gas Tungsten Arc Welding (GTAW). The welding is on V-groove using automatic GTAW technique from the inside of the casing. The penetration bead (the back side of the weld) therefore needs to be finished with machining tools to inspect the qualification of the welding. This paper summarizes the development of the special purpose micro machines which are installed inside the narrow gap being provided between the RIP casing and the RPV (skirt) to finish the penetration bead. (author)

  18. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  19. Transport Phenomena in Thin Rotating Liquid Films Including: Nucleate Boiling

    Science.gov (United States)

    Faghri, Amir

    2005-01-01

    In this grant, experimental, numerical and analytical studies of heat transfer in a thin liquid film flowing over a rotating disk have been conducted. Heat transfer coefficients were measured experimentally in a rotating disk heat transfer apparatus where the disk was heated from below with electrical resistance heaters. The heat transfer measurements were supplemented by experimental characterization of the liquid film thickness using a novel laser based technique. The heat transfer measurements show that the disk rotation plays an important role on enhancement of heat transfer primarily through the thinning of the liquid film. Experiments covered both momentum and rotation dominated regimes of the flow and heat transfer in this apparatus. Heat transfer measurements have been extended to include evaporation and nucleate boiling and these experiments are continuing in our laboratory. Empirical correlations have also been developed to provide useful information for design of compact high efficiency heat transfer devices. The experimental work has been supplemented by numerical and analytical analyses of the same problem. Both numerical and analytical results have been found to agree reasonably well with the experimental results on liquid film thickness and heat transfer Coefficients/Nusselt numbers. The numerical simulations include the free surface liquid film flow and heat transfer under disk rotation including the conjugate effects. The analytical analysis utilizes an integral boundary layer approach from which

  20. Atmospheric Pressure Effects on Cryogenic Storage Tank Boil-Off

    Science.gov (United States)

    Sass, J. P.; Frontier, C. R.

    2007-01-01

    The Cryogenics Test Laboratory (CTL) at the Kennedy Space Center (KSC) routinely utilizes cryostat test hardware to evaluate comparative and absolute thermal conductivities of a wide array of insulation systems. The test method is based on measurement of the flow rate of gas evolved due to evaporative boil-off of a cryogenic liquid. The gas flow rate typically stabilizes after a period of a couple of hours to a couple of days, depending upon the test setup. The stable flow rate value is then used to calculate the thermal conductivity for the insulation system being tested. The latest set of identical cryostats, 1,000-L spherical tanks, exhibited different behavior. On a macro level, the flow rate did stabilize after a couple of days; however the stable flow rate was oscillatory with peak to peak amplitude of up to 25 percent of the nominal value. The period of the oscillation was consistently 12 hours. The source of the oscillation has been traced to variations in atmospheric pressure due to atmospheric tides similar to oceanic tides. This paper will present analysis of this phenomenon, including a calculation that explains why other cryostats are not affected by it.

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

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

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

  2. Model for boiling and dryout in particle beds

    International Nuclear Information System (INIS)

    Over the last ten years experiments and modeling of dryout in particle beds have produced over fifty papers. Considering only volume-heated beds, over 250 dryout measurements have been made, and are listed in this work. In addition, fifteen models to predict dryout have been produced and are discussed. A model is developed in this report for one-dimensional boiling and dryout in a porous medium. It is based on conservation laws for mass, momentum, and energy. The initial coupled differential equations are reduced to a single first-order differential equation with an algebraic equation for the upper boundary condition. The model includes the effects of both laminar and turbulent flow, two-phase friction, and capillary force. The boundary condition at the bed bottom includes the possibility of inflowing liquid and either an adiabatic or a bottom-cooled support structure. The top of the bed may be either channeled or subcooled. In the first case the channel length and the saturation at the base of the channels are predicted. In the latter case, a criterion for penetration of the subcooled zone by channels is obtained

  3. Critical heat flux of an impinging water jet on a heated surface with boiling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [Andong Institute of Informaion Technology, Andong (Korea); Kim, H.D. [Andong National University, Andong (Korea); Choi, K.W. [Incheon University, Incheon (Korea)

    2000-04-01

    The purpose of this paper is to investigate a critical heat flux(CHF) during forced convective subcooled and saturated boiling in free water jet system impinged on a rectangular heated surface. The surface is supplied with subcooled or saturated water through a rectangular jet. Experimental parameters studied are a width of heated surface, a height of supplementary water and a degree of subcooling. Incipient boiling point is observed in the temperature of 6{approx}8 deg.C of superheat of test specimen. CHF depends on jet velocity for various boiling-involved coolant system. CHF also is proportional to the nozzle exit velocity to the power of n, where n is 0.55 and 0.8 for subcooled and saturated boiling, respectively. CHF is enhanced with a higher jet velocity, higher degree of subcooling and smaller width of a heated surface. (author). 18 refs., 13 figs., 1 tab.

  4. Boiling crisis as inhibition of bubble detachment by the vapor recoil force

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Yves

    2016-01-01

    Boiling crisis is a transition between nucleate and film boiling. In this communication we present a physical model of the boiling crisis based on the vapor recoil effect. Our numerical simulations of the thermally controlled bubble growth at high heat fluxes show how the bubble begins to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes the vapor spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further bubble spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. Under such conditions, we observed an increase of the apparent contact angle and spreading of the dry spot under the bubble, thus confirming our model of the boiling crisis.

  5. An Experimental Study of Boiling in Reduced and Zero Gravity Fields

    Science.gov (United States)

    Usiskin, C. M.; Siegel, R.

    1961-01-01

    A pool boiling apparatus was mounted on a counterweighted platform which could be dropped a distance of nine feet. By varying the size of the counterweight, the effective gravity field on the equipment was adjusted between zero and unity. A study of boiling burnout in water indicated that a variation in the critical heat flux according to the one quarter power of gravity was reasonable. A consideration of the transient burnout process was necessary in order to properly interpret the data. A photographic study of nucleate boiling showed how the velocity of freely rising vapor bubbles decreased as gravity was reduced. The bubble diameters at the time of breakoff from the heated surface were found to vary inversely as gravity to the 1/3.5 power. Motion pictures were taken to illustrate both nucleate and film boiling in the low gravity range.

  6. Experimental Investigation on Pool Boiling Heat Transfer With Ammonium Dodecyl Sulfate

    Directory of Open Access Journals (Sweden)

    Mr.P. Atcha Rao

    2015-11-01

    Full Text Available We have so many applications related to Pool Boiling. The Pool Boiling is mostly useful in arid areas to produce drinking water from impure water like sea water by distillation process. It is very difficult to distill the only water which having high surface tension. The surface tension is important factor to affect heat transfer enhancement in pool boiling. By reducing the surface tension we can increase the heat transfer rate in pool boiling. From so many years we are using surfactants domestically. It is proven previously by experiments that the addition of little amount of surfactant reduces the surface tension and increase the rate of heat transfer. There are different groups of surfactants. From those I‟m conducting experimentation with anionic surfactant Ammonium Dodecyl Sulfate (ADS, which is most human friendly and three times best soluble than Sodium Dodecyl Sulfate, to test the heat transfer enhancement.

  7. Boiling crisis as inhibition of bubble detachment by the vapor recoil force

    International Nuclear Information System (INIS)

    Boiling crisis is a transition between nucleate and film boiling. In this communication we present a physical model of the boiling crisis based on the vapor recoil effect. Our numerical simulations of the thermally controlled bubble growth at high heat fluxes show how the bubble begins to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes the vapor spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further bubble spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. Under such conditions, we observed an increase of the apparent contact angle and spreading of the dry spot under the bubble, thus confirming our model of the boiling crisis. (authors)

  8. Sensory quality and appropriateness of raw and boiled Jerusalem artichoke tubers (Helianthus tuberosus L.)

    DEFF Research Database (Denmark)

    Bach, Vibe; Kidmose, Ulla; Thybo, Anette;

    2013-01-01

    BACKGROUND: The aim of the present study was to investigate the sensory attributes, dry matter and sugar content of five varieties of Jerusalem artichoke tubers and their relation to the appropriateness of the tubers for raw and boiled preparation. RESULTS: Sensory evaluation of raw and boiled...... Jerusalem artichoke tubers was performed by a trained sensory panel and a semi-trained consumer panel of 49 participants, who also evaluated the appropriateness of the tubers for raw and boiled preparation. The appropriateness of raw Jerusalem artichoke tubers was related to Jerusalem artichoke flavour......, green nut flavour, sweetness and colour intensity, whereas the appropriateness of boiled tubers was related to celeriac aroma, sweet aroma, sweetness and colour intensity. In both preparations the variety Dwarf stood out from the others by being the least appropriate tuber. CONCLUSION: A few sensory...

  9. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact Regarding an Exemption Request AGENCY: Nuclear Regulatory Commission. ACTION: Environmental assessment and... Waste Management and Environmental Protection, Office of Federal and State Materials and...

  10. CFD Simulations and Experimental Verification on Nucleate Pool Boiling of Liquid Nitrogen

    Science.gov (United States)

    Xiaobin, Zhang; Wei, Xiong; Jianye, Chen; Yuchen, Wang; Tang, K.

    To explore the mechanism of nucleate pool boiling of cryogenic fluids, an experimental apparatus was built to conduct a visualization study and verify the CFD boiling model. Apart from the general measurements of the super-heat and heat flux, the influences of super-heat on bubble departure diameters were specially analyzed. Based on the observations, the whole nucleate boiling process from bubble formation to departure from the heated wall can be divided into three stages: low heat flux stage; transitional stage; fully developed nucleate boiling (FDNB) stage. CFD simulations with several existing correlations and the attained values from the experiments for the bubble diameter were finally conducted, and the results fitted well with the present experimental data.

  11. Heat Transfer in Nucleate Pool Boiling of Binary and Ternary Refrigerant Mixtures

    Institute of Scientific and Technical Information of China (English)

    赵耀华; 刁彦华; 鹤田隆治; 西川日出男

    2004-01-01

    Heat transfer coefficients in nucleate pool boiling were measured on a horizontal copper surface for refrigerants, HFC-134a, HFC-32, and HFC-125, their binary and ternary mixtures under saturated conditions at 0.9MPa. Compared to pure components, both binary and ternary mixtures showed lower heat transfer coefficients.This deterioration was more pronounced as heat flux was increased. Experimental data were compared with some empirical and semi-empirical correlations available in literature. For binary mixture, the accuracy of the correlations varied considerably with mixtures and the heat flux. Experimental data for HFC-32/134a/125 were also compared with available correlated equation obtained by Thome. For ternary mixture, the boiling range of binary mixture composed by the pure fluids with the lowest and the medium boiling points, and their concentration difference had important effects on boiling heat transfer coefficients.

  12. Pool boiling of dielectric liquids on porous graphite and extended copper surfaces

    Science.gov (United States)

    Parker, Jack L.

    This work investigated pool boiling of the dielectric liquids HFE-7100 and FC-72 on plane copper and porous graphite and on copper surfaces with corner pins. The work investigated the effects of surface orientation and liquid subcooling and, for the copper surfaces with corner pins, the effect of surface roughness. In addition, investigations were made studying the heat transfer by natural convection and nucleate boiling, as well as the effects of liquid subcooling (up to 30 K) and surface inclination (0°--upward facing, to 180°--downward facing) on nucleate boiling heat transfer and Critical Heat Flux (CHF). The results are applicable to direct immersion cooling by nucleate boiling of high power computer chips dissipating 50 - 100 W/cm2 while maintaining the junction temperature for the chips below the recommended values (˜85 °C). Pool boiling experiments are performed with degassed HFE-7100 and FC-72 liquids using uniformly heated 10 x 10 mm porous graphite and copper surfaces with corner pins. The measured footprint temperatures and thermal power removed from the surfaces are used to construct the pool boiling curves and determine the critical heat flux and corresponding surface superheat. Results are compared with those obtained on plane copper of same heated footprint area. The obtained CHF values are also compared with those reported in the open literature for plane, micro-porous, and macro-structured surfaces. Digital photographs and video are obtained to help explain and interpret the results. For the first time, natural convection correlations for dielectric liquids on plane, porous, and copper with corner pins developed. These correlations are important to electronic cooling in the stand-by mode when the heat dissipation by the chips is only a few watts. Results show that the power removed by natural convection from surfaces with corner pins is 67% more than from plane Si and Cu surfaces at the same surface superheat. Using porous graphite and copper

  13. Experimental study on a new solar boiling water system with holistic track solar funnel concentrator

    International Nuclear Information System (INIS)

    A new solar boiling water system with conventional vacuum-tube solar collector as primary heater and the holistic solar funnel concentrator as secondary heater had been designed. In this paper, the system was measured out door and its performance was analyzed. The configuration and operation principle of the system are described. Variations of the boiled water yield, the temperature of the stove and the solar irradiance with local time have been measured. Main factors affecting the system performance have been analyzed. The experimental results indicate that the system produced large amount of boiled water. And the performance of the system has been found closely related to the solar radiance. When the solar radiance is above 600 W/m2, the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  14. X ray observations of boiling sodium in a reflux-pool-boiler solar receiver

    Science.gov (United States)

    Moreno, J. B.; Stoker, G. C.; Thompson, K. R.

    1992-01-01

    X ray observations of boiling sodium in a 75-kW sub t reflux-pool-boiler solar receiver operating at up to 800 C were carried out. Both cinematographic and quantitative observations were made. From the cinematography, the pool free surface was observed before and during the start of boiling. During boiling, the free surface rose out of the field of view, and chaotic motion was observed. From the quantitative observations, void fraction in pencil-like probe volumes was inferred, using a linear array of detectors. Useful data were obtained from three of the eight probe volumes. Information from the other volumes was masked by scattered radiation. During boiling, time-averaged void fractions ranged from 0.6 to 0.8. During hot restarts, void fractions near unity occurred and persisted for up to 1/2 second.

  15. Observation of high heat flux boiling structures in a horizontal pool by a total reflection technique

    International Nuclear Information System (INIS)

    The experiments were carried out for a horizontal pool boiling of saturated water using a transparent ITO heating surface. Details of boiling structure near the heated surface have been clearly observed by applying the total reflection and diagonal view techniques in a synchronized manner. Mechanisms for the bubble coalescence and dry area expansion processes were clearly identified. The base of the large massive bubble was mostly dry with some trapped liquid. The appearance of this large dry area at high heat flux close to CHF was basically resulted from the multiple steps of bubble coalescences which occur while the bubbles are growing, attached to the boiling surface not before they depart from the boiling surface. The thin liquid layer with distributed vapor stems was not observed under the large massive bubble. (author)

  16. Transient CHF enhancement of saturated pool boiling of water using a honeycomb porous media

    International Nuclear Information System (INIS)

    Several studies have been performed to make clear the transient boiling heat transfer during the exponential heat generation which is occurred in reactivity accident of a nuclear reactor. These researches have been focused on the mechanism of the phenomena mainly, not on the enhancement of the transient boiling heat transfer. In a previous study, we proposed a method of CHF enhancement under steady-state conditions using honeycomb porous plate. The CHF was shown experimentally to be enhanced to more than twice that of a plain surface using honeycomb porous plate. The enhancement is considered to result from the capillary supply of liquid onto the heated surface and the release of generated vapor through the channels. In the present paper, enhancement of the transient critical heat flux in pool boiling by the attachment of a honeycomb-structured porous plate on a heated wire is investigated experimentally using water under saturated boiling conditions. (author)

  17. Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

    Science.gov (United States)

    Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

    2015-10-01

    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

  18. Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

    Science.gov (United States)

    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

  19. Flow boiling of R245fa in vertical small metallic tubes

    OpenAIRE

    Pike-Wilson, Emily Alexandra

    2014-01-01

    This thesis was submitted for the degree of Doctor of Philosophy and awarded by Brunel University London The research presented is part of a larger study, dedicated to investigating flow boiling in small to microchannels. The test facility, originally designed by Huo (2005) and since used by Chen (2006) and Mahmoud (2011), has been used to investigate flow boiling of R134a across a range of channel diameters and both seamless cold drawn and welded channels. These previous studies concluded...

  20. Enhanced convective and film boiling heat transfer by surface gas injection

    Energy Technology Data Exchange (ETDEWEB)

    Duignan, M.R.; Greene, G.A. [Brookhaven National Lab., Upton, NY (United States); Irvine, T.F., Jr. [State Univ. of New York, Stony Brook, NY (United States). Dept. of Mechanical Engineering

    1992-04-01

    Heat transfer measurements were made for stable film boiling of water over a horizontal, flat stainless steel plate from the minimum film boiling point temperature, T{sub SURFACE} {approximately}500K, to T{sub SURFACE} {approximately}950K. The pressure at the plate was approximately 1 atmosphere and the temperature of the water pool was maintained at saturation. The data were compared to the Berenson film-boiling model, which was developed for minimum film-boiling-point conditions. The model accurately represented the data near the minimum film-boiling point and at the highest temperatures measured, as long it was corrected for the heat transferred by radiation. On the average, the experimental data lay within {plus_minus}7% of the model. Measurements of heat transfer were made without film boiling for nitrogen jetting into an overlying pool of water from nine 1-mm- diameter holes, drilled in the heat transfer plate. The heat flux was maintained constant at approximately 26.4 kW/m{sup 2}. For water-pool heights of less than 6cm the heat transfer coefficient deceased linearly with a decrease in heights. Above 6cm the heat transfer coefficient was unaffected. For the entire range of gas velocities measured [0 to 8.5 cm/s], the magnitude of the magnitude of the heat transfer coefficient only changed by approximately 20%. The heat transfer data bound the Konsetov model for turbulent pool heat transfer which was developed for vertical heat transfer surfaces. This agreement suggests that surface orientation may not be important when the gas jets do not locally affect the surface heat transfer. Finally, a database was developed for heat transfer from the plate with both film boiling and gas jetting occurring simultaneously, in a pool of water maintained at its saturation temperature. The effect of passing nitrogen through established film boiling is to increase the heat transfer from that surface. 60 refs.

  1. Enhanced convective and film boiling heat transfer by surface gas injection

    Energy Technology Data Exchange (ETDEWEB)

    Duignan, M.R.; Greene, G.A. (Brookhaven National Lab., Upton, NY (United States)); Irvine, T.F., Jr. (State Univ. of New York, Stony Brook, NY (United States). Dept. of Mechanical Engineering)

    1992-04-01

    Heat transfer measurements were made for stable film boiling of water over a horizontal, flat stainless steel plate from the minimum film boiling point temperature, T{sub SURFACE} {approximately}500K, to T{sub SURFACE} {approximately}950K. The pressure at the plate was approximately 1 atmosphere and the temperature of the water pool was maintained at saturation. The data were compared to the Berenson film-boiling model, which was developed for minimum film-boiling-point conditions. The model accurately represented the data near the minimum film-boiling point and at the highest temperatures measured, as long it was corrected for the heat transferred by radiation. On the average, the experimental data lay within {plus minus}7% of the model. Measurements of heat transfer were made without film boiling for nitrogen jetting into an overlying pool of water from nine 1-mm- diameter holes, drilled in the heat transfer plate. The heat flux was maintained constant at approximately 26.4 kW/m{sup 2}. For water-pool heights of less than 6cm the heat transfer coefficient deceased linearly with a decrease in heights. Above 6cm the heat transfer coefficient was unaffected. For the entire range of gas velocities measured (0 to 8.5 cm/s), the magnitude of the magnitude of the heat transfer coefficient only changed by approximately 20%. The heat transfer data bound the Konsetov model for turbulent pool heat transfer which was developed for vertical heat transfer surfaces. This agreement suggests that surface orientation may not be important when the gas jets do not locally affect the surface heat transfer. Finally, a database was developed for heat transfer from the plate with both film boiling and gas jetting occurring simultaneously, in a pool of water maintained at its saturation temperature. The effect of passing nitrogen through established film boiling is to increase the heat transfer from that surface. 60 refs.

  2. Physical concept and calculation of boiling point in a pulsating heat pipe

    OpenAIRE

    Naumova A. N.; Kravets V. Yu.; Nikolaenko Yu. E.

    2014-01-01

    LED development is accompanied by the need to ensure a constructive solution for the thermal conditions problem. For this purpose one can use pulsating heat pipes (PHP), that operate more efficiently after the start of heat carrier boiling. This article describes the physical representation and formula that allows determining the boiling point, which is a lower bound of the PHP effective operating range. It is shown that the main factors influencing the required heat flow are driving capillar...

  3. Combined effect of electric field and surface modification on pool boiling of R-123

    OpenAIRE

    Ahmad, Syed Waqas

    2012-01-01

    This thesis was submitted for the degree of Doctor of Philosophy and awarded by Brunel University. The effect of surface modification and high intensity electric field (uniform and non – uniform) acting separately or in combination on pool boiling of R-123 is presented in this thesis. The effect of surface modification was investigated on saturated pool boiling of R-123 for five horizontal copper surfaces modified by different treatments, namely: an emery polished surface, a fine sandblast...

  4. Prediction of bubble departure in forced convection boiling: a mechanistic model

    OpenAIRE

    Colombo, M; Fairweather, M.

    2015-01-01

    In the context of computational fluid dynamic simulations of boiling flows using time-averaged Eulerian multi-phase approaches, the many sub-models required to describe such a complex phenomena are of particular importance. Of interest here, wall boiling requires calculation of the contribution of evaporation to global heat transfer, which in turn relies on determination of the active nucleation site density, bubble departure diameter and frequency of bubble departure. In this paper, an impro...

  5. Heat Transfer of Single and Binary Systems inPool Boiling

    OpenAIRE

    Abbas J. Sultan; Balasim A. Abid

    2010-01-01

    The present research focuses on the study of the effect of mass transfer resistance on the rate of heat transfer in pool boiling. The nucleate pool boiling heat transfer coefficients for binary mixtures (ethanol-n-butanol, acetone-n-butanol, acetone-ethanol, hexane-benzene, hexane-heptane, and methanol-water) were measured at different concentrations of the more volatile components. The systems chosen covered a wide range of mixture behaviors.The experimental set up for the present investigat...

  6. Boiling liquid expanding vapour explosion in CO2 small scale experiments

    OpenAIRE

    Bjerketvedt, Dag; Egeberg, Kjersti; Ke, Wei; Gaathaug, Andre Vagner; Vågsæther, Knut; Nilsen, S. H.

    2011-01-01

    Carbon capture and storage systems require handling large volumes of high pressure CO2. Having thorough knowledge of the related hazards is essential, as is knowing how to prevent, detect, control and mitigate accidents. This paper gives a short description of CO2 Boiling Liquid Expanding Vapour Explosions (BLEVEs) and presents results from preliminary, small scale experiments with CO2 BLEVEs. The mechanism of superheated liquid CO2 boiling is not fully understood. Analogies can be made betwe...

  7. Confinement by Carbon Nanotubes Drastically Alters the Boiling and Critical Behavior of Water Droplets

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Victor V.; Prezhdo, Oleg V.

    2012-01-01

    Vapor pressure grows rapidly above the boiling temperature, and past the critical point liquid droplets disintegrate. Our atomistic simulations show that this sequence of events is reversed inside carbon nanotubes (CNT). Droplets disintegrate first and at low temperature, while pressure remains small. The droplet disintegration temperature is independent of the CNT diameter. In contrast, depending on CNT diameter, a temperature that is much higher than the bulk boiling temperature is required...

  8. Boiling of Binary Zeotropic Blends in the Plate Heat Exchanger of the Heat Pump

    Directory of Open Access Journals (Sweden)

    Mezentseva Nadezhda N.

    2016-01-01

    Full Text Available In this paper, we consider the process of boiling in the evaporator of the heat pump. Zeotropic binary refrigerants R32/R152a (30/70% and R32/R134a (30/70% are used as working medium. Calculations are made for brazed plate heat exchanger during boiling of zeotropic blend refrigerants with account of peculiarities of this process. Results of calculation of the heat transfer coefficient for zeotropic blends are given.

  9. Transient boiling and void formation during postulated reactivity-initiated accident in BWR: Experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (author)

  10. Experimental Study of Pool Boiling Heat Transfer Enhancement with R123 under Non Uniform Electric Field

    Directory of Open Access Journals (Sweden)

    Hongling Yu

    2013-02-01

    Full Text Available Experimental investigations are carried out to study the effect of a non uniform electric field on the boiling heat transfer. The study has found that the heat transfer coefficient increases as the electric field strength increases. Enhanced coefficient decreases with heat flux increases and finally reaches a steady value. When the heat flux is small, high voltage has a better enhancement effect. The Onset of Nucleate Boiling (ONB undergoes a larger increase by applying a high voltage.

  11. Photographic study of bubble departure diameter in saturated pool boiling to electrolyte solutions

    OpenAIRE

    Peyghambarzadeh S.M.; Hatami A.; Ebrahimi A; Fazel Alavi S.A.

    2014-01-01

    Bubble departure diameters during saturated pool boiling to pure water and three different electrolyte solutions including NaCl, KNO3, and KCl aqueous solutions are experimentally measured. Variable heat fluxes up to 90 kW/m2 and different salt concentrations from 10.6 to 69.6 kg/m3 are applied in order to investigate their effects on the bubble size during pool boiling around the horizontal rod heater. Visual observations demonstrated that larger vapor bub...

  12. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Lecoutre, C; Chatain, D

    2016-01-01

    Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil the gas bubbles begin to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes its spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. In the experiments both in the Mir spa...

  13. Experimental study of the characteristics of pool boiling CHF enhancement using water-based magnetic fluid

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Hyuk; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2011-05-15

    Nucleate boiling is a very effective heat transfer mechanism. However, there exists a critical value of heat flux at which nucleate boiling transitions to film boiling shows very poor heat transfer behavior. Critical heat flux(CHF) is a main constraint to the design process because it can generate damages or deformations of material. There have been many efforts to improve the CHF by using nanofluids by researchers. This paper will describe the effects of magnetic fluid on CHF enhancement of pool boiling. We compared the CHF values of pool boiling experiment between magnetic fluid and other nanofluids with several volume concentrations to evaluate the degree of CHF enhancement. SEM(Scanning Electron Microscope) images were obtained to explain CHF enhancement through the effect of the deposited nanoparticles, which can change the surface wettability, during the pool boiling experiment. Lastly, Finally, in order to investigate the effect of magnetic field in the water-based magnetic fluid, magnetic field was analytically calculated by using Biot-Savart law. Using these results, we discussed the CHF enhancement of magnetite-water nanofluids in detailed

  14. Burnout in subcooled flow boiling of water. A visual experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A.; Zummo, G. [ENEA, Engineering Div., National Institute of Thermal Fluid-Dynamics, Rome (Italy); Cumo, M. [University of Rome la Sapienza, Rome (Italy)

    2000-12-01

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  15. Investigation of film boiling thermal hydraulics under FCI conditions. Results of a numerical study

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Dinh, A.T.; Nourgaliev, R.R.; Sehgal, B.R. [Div. of Nuclear Power Safety Royal Inst. of Tech. (RIT), Brinellvaegen 60, 10044 Stockholm (Sweden)

    1998-01-01

    Film boiling on the surface of a high-temperature melt jet or of a melt particle is one of key phenomena governing the physics of fuel-coolant interactions (FCIs) which may occur during the course of a severe accident in a light water reactor (LWR). A number of experimental and analytical studies have been performed, in the past, to address film boiling heat transfer and the accompanying hydrodynamic aspects. Most of the experiments have, however, been performed for temperature and heat flux conditions, which are significantly lower than the prototypic conditions. For ex-vessel FCIs, high liquid subcooling can significantly affect the FCI thermal hydraulics. Presently, there are large uncertainties in predicting natural-convection film boiling of subcooled liquids on high-temperature surfaces. In this paper, research conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning film-boiling thermal hydraulics under FCI condition is presented. Notably, the focus is placed on the effects of (1) water subcooling, (2) high-temperature steam properties, (3) the radiation heat transfer and (4) mixing zone boiling dynamics, on the vapor film characteristics. Numerical investigations are performed using a novel CFD modeling concept named as the local-homogeneous-slip model (LHSM). Results of the analytical and numerical studies are discussed with respect to boiling dynamics under FCI conditions. (author)

  16. Burnout in subcooled flow boiling of water. A visual experimental study

    International Nuclear Information System (INIS)

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  17. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    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.

  18. Uncommon water chemistry observations in modern day boiling water reactors

    International Nuclear Information System (INIS)

    Numerous technologies have been developed to mitigate intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) materials that include hydrogen water chemistry (HWC), noble metal chemical application (NMCA) and on-line NMCA (OLNC). These are matured technologies with extensive plant operating experiences, HWC – 32 years, NMCA – 18 years and OLNC – 9 years. Over the past three decades, numerous water chemistry data, dose rate data and IGSCC mitigation data relating to these technologies have been published and presented at many international conferences. However, there are many valuable and critical water chemistry and dose rate data that have gone unnoticed and unreported. The purpose of this paper is to highlight some of the uncommon water chemistry and dose rate experiences that reveal valuable information on the performance and durability of NMCA and OLNC technologies. Data will be presented, that have hitherto been unseen in public domain, from the lead OLNC plant in Switzerland giving reasons for some of the uncommon or overlooked water chemistry observations. They include, decreasing reactor water platinum concentration with each successive OLNC application, lack of increase in reactor water activation products in later applications, gradual disappearance of main steam line radiation (MSLR) monitor response decrease, Curium and Au-199 release during OLNC applications, rapid increase in reactor water clean-up conductivity, and Iodine, Mo-99 and Tc-99m spiking when hydrogen is interrupted and brought back to service, and main steam and reactor water conductivity spiking when clean-up beds or condensate demineralizers are changed. All these observations give valuable information on the success of OLNC applications and also signal the presence of sufficient noble metal on in-reactor surfaces from the long term durability and effectiveness stand point. Some of these observations can be used as secondary parameters, if and when a primary

  19. Calculations of the effect of boiling water on bitumen production

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.; Kantzas, A. [Calgary Univ., AB (Canada). Dept. of Chemical and Petroleum Engineering]|[Calgary Univ., AB (Canada). Tomographic Imaging and Porous Media Laboratory; McGee, B. [E-T Energy Limited, Calgary, AB (Canada)

    2006-07-01

    Alberta's vast resources of heavy oil and bitumen are playing an increasing role as a main resource for crude oil. Thermal recovery methods for heavy oil and bitumen include steam injection and steam flooding in which thermal energy is given to the oil to reduce its viscosity and allow it to flow towards a production spot. A viable alternative to steam injection is the electromagnetic heating method for heavy oil and bitumen reservoirs. Electromagnetic heating transfers heat to heavy oil reservoirs based on electromagnetic energy and can be used in situations where steam injection may not work well. The process can also be used to preheat the reservoir before steam injection. This study examined the possible displacement mechanisms of such processes with particular focus on the physics of boiling water in porous media as a potential displacement agent for heavy oil and bitumen. It is very possible that water could vaporize while being electrically heated and the vaporized water could push more heavy oil or bitumen out of reservoir. As such, higher oil recovery could be expected due to water vaporization. The role of water vaporization during electrical heating process was examined and a methodology to estimate the magnitude of incremental oil recovery was developed based on simple conceptual models with numerical simulators and illustrative experiments. The primary contributors of this process appear to be a combination of drainage, imbibition, viscosity reduction and gas expansion. The study showed that the expansion of water into steam could very efficiently flush oil out of pore spaces. It was concluded that water vaporization inside the reservoir can be an additional driving force for heavy oil or bitumen production, and that this alternative to steam injection can offer energy savings for the recovery process. 10 refs., 4 tabs., 6 figs., 1 appendix.

  20. Experimental investigation of wall heat flux partitioning during subcooled nucleate boiling on a vertical wall

    Energy Technology Data Exchange (ETDEWEB)

    Song, Junkyu; Park, Junseok; Jung, Satbyoul; Kim, Hyungdae [Kyung Hee Univ., Youngin (Korea, Republic of)

    2013-10-15

    This study aims to obtain the spatially and temporally synchronized experimental data of liquid-vapor phase and local heat flux distributions on the heated wall during subcooled nucleate boiling, to analyze the data based on the fundamental physical parameters associated with boiling. In this paper, the infrared thermometry and the total reflection techniques were spatially and temporally synchronized in during subcooled vertical plate boiling. The three fundamental heat transfer mechanisms in RPI model of nucleate boiling, evaporation, quenching and convection, were separately detected and calculated from the obtained high-resolution experimental data. The contribution of each heat removal mechanism was found to be 24 %, 39 % and 37 %, respectively, while the only quenching heat flux was dominant (∼95%) in the analyses using heat partitioning correlation of the commercial and developing computational analysis codes, including Fluent and CUPID. A number of experimental investigations have been conducted for the understanding of the exact mechanisms of subcooled flow boiling and critical heat flux (CHF). Bang et al. conducted a visualization study of CHF and found evidence of a liquid layer beneath the large vapor mushroom. Geradi et al. measured time- and space-resolved temperature distribution on bubble nucleation and boiling heat transfer on an ITO-film-coated glass heater by means of the synchronized high-speed video and IR thermometry. There also have been many numerical simulation studies on flow boiling heat transfer. Yun et al. performed the studies to improve the prediction accuracy of subcooled flow boiling heat transfer. However, our understanding of the physical mechanism is still not enough to accurately model boiling heat transfer phenomena with application to the high-fidelity computational thermal-hydraulic analysis code. As nucleate boiling heat transfer and CHF occur along with complex mutual interactions of two-phase flow and transient wall heat

  1. Numerical simulation of boiling and two-phase flow in PCCT of PAFS

    International Nuclear Information System (INIS)

    The Passive Auxiliary Feedwater System (PAFS) is one of the passive safety features adopted in the Advanced Power Reactor Plus (APR+) and replaces the conventional active auxiliary feed-water system of the APR1400 by introducing a natural driving force mechanism (Fig. 1). It has a function of removing the decay heat and the residual heat. The PAFS is composed of two independent trains (each 100% capability) to satisfy the single failure criterion. In each train, one Passive Condensation Heat exchangers (PCHX) is installed inside the Passive Condensation Cooling Tank (PCCT) as shown in Fig. 2. The PAFS is designed to have a capability of operating without AC power or operator action for a minimum of 8 hours in 5 minutes after reactor trip and to ensure a subsequent RCS cooldown for 8 hours to shutdown cooling entry conditions. The steam generated in the steam generator is delivered to the PCHX, and is condensed in the tube inside of PCHX. The condensate is conveyed to the economizer of the steam generator by a natural circulation system. The PCCT provides the heat sink for the PCHX. The heat addition from the PCHX initially increases the water temperature. As time passes, the water reaches saturation temperature, and then begins to boil. The steam generated in the PCCT is discharged to atmosphere. The cooling water in the PCCT circulates due to the rising steam bubbles. In process of time, the water level gets lower gradually. The purpose of this study is to offer detailed information on thermal hydrodynamic phenomena in the PCCT of the PAFS. The thermal hydrodynamic phenomena in the PCCT are simulated using a computational fluid dynamics (CFD) technique

  2. Numerical simulation of boiling and two-phase flow in PCCT of PAFS

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Choeng Ryul [ELSOLTEC Inc., Suwon (Korea, Republic of); Kim, Kwang Chu; Seong, Ho Je; Kim, Tae Young [KEPCO E and C, Yongin (Korea, Republic of)

    2011-10-15

    The Passive Auxiliary Feedwater System (PAFS) is one of the passive safety features adopted in the Advanced Power Reactor Plus (APR+) and replaces the conventional active auxiliary feed-water system of the APR1400 by introducing a natural driving force mechanism (Fig. 1). It has a function of removing the decay heat and the residual heat. The PAFS is composed of two independent trains (each 100% capability) to satisfy the single failure criterion. In each train, one Passive Condensation Heat exchangers (PCHX) is installed inside the Passive Condensation Cooling Tank (PCCT) as shown in Fig. 2. The PAFS is designed to have a capability of operating without AC power or operator action for a minimum of 8 hours in 5 minutes after reactor trip and to ensure a subsequent RCS cooldown for 8 hours to shutdown cooling entry conditions. The steam generated in the steam generator is delivered to the PCHX, and is condensed in the tube inside of PCHX. The condensate is conveyed to the economizer of the steam generator by a natural circulation system. The PCCT provides the heat sink for the PCHX. The heat addition from the PCHX initially increases the water temperature. As time passes, the water reaches saturation temperature, and then begins to boil. The steam generated in the PCCT is discharged to atmosphere. The cooling water in the PCCT circulates due to the rising steam bubbles. In process of time, the water level gets lower gradually. The purpose of this study is to offer detailed information on thermal hydrodynamic phenomena in the PCCT of the PAFS. The thermal hydrodynamic phenomena in the PCCT are simulated using a computational fluid dynamics (CFD) technique

  3. Boiling behavior of sodium-potassium alloy in a bench-scale solar receiver

    Science.gov (United States)

    Moreno, J. B.; Andraka, C. E.; Moss, T. A.

    During 1989-90, a 75-kW(sub t) sodium reflux pool-boiler solar receiver was successfully demonstrated at Sandia National Laboratories. Significant features of this receiver include the following: (1) boiling sodium as the heat transfer medium, and (2) electric-discharge-machined (EDM) cavities as artificial nucleation sites to stabilize boiling. Since this first demonstration, design of a second-generation pool-boiler receiver that will bring the concept closer to commercialization has begun. For long life, the new receiver uses Haynes Alloy 230. For increased safety factors against film boiling and flooding, it has a refined shape and somewhat larger dimensions. To eliminate the need for trace heating, the receiver will boil the sodium-potassium alloy NaK-78 instead of sodium. To reduce manufacturing costs, it will use one of a number of alternatives to EDM cavities for stabilization of boiling. To control incipient-boiling superheats, especially during hot restarts, it will contain a small amount of inert gas. Before the new receiver design could be finalized, bench-scale tests of some of the proposed changes were necessary. A series of bench-scale pool boilers were built from Haynes Alloy 230 and filled with NaK-78. Various boiling-stabilizer candidates were incorporated into them, including laser-drilled cavities and a number of different sintered-powder-metal coatings. These bench-scale pool boilers have been operated at temperatures up to 750 C, heated by quartz lamps with incident radiant fluxes up to 95 W/sq cm. The effects of various orientations and added gases have been studied. Results of these studies are presented.

  4. Experimental Study and Heat Transfer Analysis on the Boiling of Saturated Liquid Nitrogen under Transient Pulsed Laser Irradiation

    Institute of Scientific and Technical Information of China (English)

    Zhaoyi DONG; Xiulan HUAI

    2005-01-01

    The boiling behavior of the liquid nitrogen (LN2) under the transient high heat flux urgently needs to be researched systematically. In this paper, the high power short pulse duration laser was used to heat the saturated LN2 rapidly, and the high-speed photography aided by the spark light system was employed to take series of photos which displayed the process of LN2's boiling behavior under such conditions. Also, a special temperature measuring system was applied to record the temperature variation of the heating surface. The experiments indicated that an explosive boiling happened within LN2 by the laser heating, and a conventional boiling followed up after the newly-defined changeover time. By analyzing the temperature variation of the heating surface, it is found that the latent heat released by the crack of the bubbles in the bubble cluster induced by the explosive boiling is an important factor that greatly influences the boiling heat transfer mechanism.

  5. Boiling transition and the possibility of spontaneous nucleation under high subcooling and high mass flux density flow in a tube

    International Nuclear Information System (INIS)

    Boiling transition and inverted annular heat transfer for R-113 have been investigated experimentally in a horizontal tube of 1.2 X 10/sup -3/ meter inner diameter with heating length over inner diameter ratio of 50. Experiments cover a high mass flux density range, a high local subcooling range and a wide local pressure range. Heat transfer characteristics were obtained by using heat flux control steady-state apparatus. Film boiling treated here is limited to the case of inverted annular heat transfer with very thin vapor film, on the order of 10/sup -6/ meter. Moreover, film boiling region is always limited to a certain downstream part, since the system has a pressure gradient along the flow direction. Discussions are presented on the parametric trends of boiling heat transfer characteristic curves and characteristic points. The possible existence is suggested of a spontaneous nucleation control surface boiling phenomena. And boiling transition heat flux and inverted annular heat transfer were correlated

  6. Non linear dynamics of boiling water reactor dynamical system

    International Nuclear Information System (INIS)

    The fifth order phenomenological model of March-Leuba for boiling water reactors include the point reactor kinetics equations for neutron balance and effective delayed neutron precursor groups with one node representation of the heat transfer process and channel thermal hydraulics. This nonlinear mathematical model consists five coupled nonlinear ordinary differential equations. The reactivity feedback (void coefficient of reactivity as well as the fuel temperature coefficient of reactivity), heat transfer process and momentum balance are major reasons for the appearance of nonlinearity in this dynamical system. The linear stability of a dynamical system with the existence of nonlinearity cannot predict a true picture of the stability characteristics of dynamical system; hence nonlinear stability analyses become an essential part to predict the global stable region on the stability map. The linear stable region is analyzed by the eigenvalues. In this stable region all the eigenvalues have negative real parts, but when pair of one of the complex eigenvalues passes transversely through imaginary axis, the dynamical system loses or gain its stability via a Hopf bifurcation and limit cycles emerges from the tip. The study of eigenvalues can predict a few bifurcations. The first Lyapunov coefficient and normal form coefficients can be used for the detection of other bifurcations in the systems. Stable or unstable limit cycles excite from these Hopf points. These limits cycles gains or loses their stability via limit point bifurcation of cycles, period doubling bifurcation of cycles and Neimark-Sacker bifurcation of cycles when one of the parameters of the nuclear dynamical system is varied. The stability of these limit cycles can be studied by Floquet theory and Lyapunov coefficient, but the bifurcations of limit cycles can be investigated only by critical Floquet multiplier which is basically the eigenvalue of the monodromy matrices. The cascade of period doubling

  7. Nucleate boiling incipience over metallic/non-metallic surfaces

    Science.gov (United States)

    Petralanda, Naiara

    /vapor contact angle. Based on measured values of the chemical potential at incipience, the wall superheat at incipience for heterogeneous boiling on smooth surfaces can be determined.

  8. Film boiling on spheres in single- and two-phase flows.

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C.; Theofanous, T. G.

    2000-08-29

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40 C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900 C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1 - {alpha}){sup 1/4} (with a being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multi-sphere structure on the film boiling heat transfer in single- and two-phase flows.

  9. Film boiling on spheres in single- and two-phase flows. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C.; Theofanous, T.G.

    1994-12-01

    Film boiling on spheres in single- and two-phase flows was studied experimentally and theoretically with an emphasis on establishing the film boiling heat transfer closure law, which is useful in the analysis of nuclear reactor core melt accidents. Systematic experimentation of film boiling on spheres in single-phase water flows was carried out to investigate the effects of liquid subcooling (from 0 to 40{degrees}C), liquid velocity (from 0 to 2 m/s), sphere superheat (from 200 to 900{degrees}C), sphere diameter (from 6 to 19 mm), and sphere material (stainless steel and brass) on film boiling heat transfer. Based on the experimental data a general film boiling heat transfer correlation is developed. Utilizing a two-phase laminar boundary-layer model for the unseparated front film region and a turbulent eddy model for the separated rear region, a theoretical model was developed to predict the film boiling heat transfer in all single-phase regimes. The film boiling from a sphere in two-phase flows was investigated both in upward two-phase flows (with void fraction from 0.2 to 0.65, water velocity from 0.6 to 3.2 m/s, and steam velocity from 3.0 to 9.0 m/s) and in downward two-phase flows (with void fraction from 0.7 to 0.95, water velocity from 1.9 to 6.5 m/s, and steam velocity from 1.1 to 9.0 m/s). The saturated single-phase heat transfer correlation was found to be applicable to the two-phase film boiling data by making use of the actual water velocity (water phase velocity), and an adjustment factor of (1-{alpha}){sup 1/4} (with {alpha} being the void fraction) for downward flow case only. Slight adjustments of the Reynolds number exponents in the correlation provided an even better interpretation of the two-phase data. Preliminary experiments were also conducted to address the influences of multisphere structure on the film boiling heat transfer in single- and two-phase flows.

  10. Burnout in a high heat flux boiling system with forced supply of liquid through a plane jet

    International Nuclear Information System (INIS)

    As for pool boiling, the non-dimensional formula for the burnout heat flux of a simple, basic boiling system has been obtained. On the other hand, in forced convection boiling, the studies on the burnout in forced flow boiling in a channel have been continued, but the derivation of a non-dimensional formula applicable generally is far away from the realization because the phenomena are too complex. Accordingly, in this study, the result of the experiment on the burnout of a boiling system to which liquid is supplied by the plane jet flowing out of a thin rectangular nozzle installed near the front edge of a rectangular heating surface is reported. The experimental apparatus is described, and the experiment was carried out in the ranges of two jet thicknesses at the nozzle outlet, two incident angles of jet and from 1.5 to 15 m/s of jet velocity. Burnout occurs under the situation of sufficiently developed nuclear boiling. A part of the liquid supplied from a plane jet is blown apart by the vapor blowing out of the nuclear boiling liquid layer covering the heating surface in the nuclear boiling with sufficiently developed high heat flux. However, the nuclear boiling liquid layer itself continues to exist on the heating surface till burnout occurs. Only the entering velocity of the plane jet affects burnout heat flux. (Kako, I.)

  11. Evidence for increasing severity of community-onset boils and abscesses in UK General Practice.

    Science.gov (United States)

    Shallcross, L J; Hayward, A C; Johnson, A M; Petersen, I

    2015-08-01

    In England, hospital admissions for severe staphylococcal boils and abscesses trebled between 1989 and 2004. We investigated this trend using routine data from primary and secondary care. We used The Health Improvement Network (THIN), a large primary-care database and national data on hospital admissions from Hospital Episode Statistics (HES). Time trends in the incidence of primary-care consultations for boils and abscesses were estimated for 1995-2010. HES data were used to calculate age-standardized hospital admission rates for boils, abscesses and cellulitis. The incidence of boil or abscess was 450 [95% confidence interval (CI) 447-452] per 100 000 person-years and increased slightly over the study period (incidence rate ratio 1·005, 95% CI 1·004-1·007). The rate of repeat consultation for a boil or abscess increased from 66 (95% CI 59-73) per 100 000 person-years in 1995 to peak at 97 (95% CI 94-101) per 100 000 person-years in 2006, remaining stable thereafter. Hospital admissions for abscesses, carbuncles, furuncles and cellulitis almost doubled, from 123 admissions per 100 000 in 1998/1999 to 236 admissions per 100 000 in 2010/2011. Rising hospitalization and recurrence rates set against a background of stable community incidence suggests increased disease severity. Patients may be experiencing more severe and recurrent staphylococcal skin disease with limited treatment options. PMID:25530161

  12. Gas bubbling-enhanced film boiling of Freon-11 on liquid metal pools. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Greene, G.A.

    1985-01-01

    In the analysis of severe core damage accidents in LWRs, a major driving force which must be considered in evaluating containment loading and fission product transport is the ex-vessel interaction between molten core debris and structural concrete. Two computer codes have been developed for this purpose, the CORCON-MOD2 model of ex-vessel, core concrete interactions and the VANESA model for aerosol generation and fission product release as a result of molten core-concrete interactions. Under a wide spectrum of reactor designs and accident sequences, it is possible for water to come into contact with the molten core debris and form a coolant pool overlying the core debris which is attacking the concrete. As the concrete decomposes, noncondensable gases are released, which bubble through the melt and across the boiling interface, affecting the liquid-liquid boiling process. Currently, the CORCON code includes the classical Berenson model for film boiling over a horizontal flat plate for this phenomenon. The objectives of this activity are to investigate the influence of transverse noncondensable gas flux on the magnitude of the stable liquid-liquid film boiling heat flux and develop a gas flux-enhanced, liquid-liquid film boiling model for incorporation into the CORCON-MOD2 computer code to replace or modify the Berenson model.

  13. Large-scale Generation of Patterned Bubble Arrays on Printed Bi-functional Boiling Surfaces

    Science.gov (United States)

    Choi, Chang-Ho; David, Michele; Gao, Zhongwei; Chang, Alvin; Allen, Marshall; Wang, Hailei; Chang, Chih-Hung

    2016-04-01

    Bubble nucleation control, growth and departure dynamics is important in understanding boiling phenomena and enhancing nucleate boiling heat transfer performance. We report a novel bi-functional heterogeneous surface structure that is capable of tuning bubble nucleation, growth and departure dynamics. For the fabrication of the surface, hydrophobic polymer dot arrays are first printed on a substrate, followed by hydrophilic ZnO nanostructure deposition via microreactor-assisted nanomaterial deposition (MAND) processing. Wettability contrast between the hydrophobic polymer dot arrays and aqueous ZnO solution allows for the fabrication of heterogeneous surfaces with distinct wettability regions. Heterogeneous surfaces with various configurations were fabricated and their bubble dynamics were examined at elevated heat flux, revealing various nucleate boiling phenomena. In particular, aligned and patterned bubbles with a tunable departure frequency and diameter were demonstrated in a boiling experiment for the first time. Taking advantage of our fabrication method, a 6 inch wafer size heterogeneous surface was prepared. Pool boiling experiments were also performed to demonstrate a heat flux enhancement up to 3X at the same surface superheat using bi-functional surfaces, compared to a bare stainless steel surface.

  14. Two-phase flow boiling in small channels: A brief review

    Indian Academy of Sciences (India)

    Madhavi V Sardeshpande; Vivek V Ranade

    2013-12-01

    Boiling flows are encountered in a wide range of industrial applications such as boilers, core and steam generators in nuclear reactors, petroleum transportation, electronic cooling and various types of chemical reactors. Many of these applications involve boiling flows in conventional channels (channel size ≥ 3 mm). The key design issues in two phase flow boiling are variation in flow regimes, occurrence of dry out condition, flow instabilities, and understanding of heat transfer coefficient and vapor quality. This paper briefly reviews published experimental and modeling work in these areas. An attempt is made to provide a perspective and to present available information on boiling in small channels in terms of channel size, flow regimes, heat transfer correlations, pressure drop, critical heat flux and film thickness. An attempt is also made to identify strengths and weaknesses of published approaches and computational models of boiling in small channels. The presented discussion and results will provide an update on the state-of-the-art and will be useful to identify and plan further research in this important area.

  15. Heat transfer behavior on small heaters during saturated pool boiling of FC-72 in microgravity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.; Mullen, J.D. [Maryland Univ., College Park, MD (United States). Dept. of Mechanical Engineering; Yaddanapudi, N. [MetaSensors, Rockville, MD (United States)

    1999-07-01

    Saturated pool boiling of FC-72 on an array of 96 heaters, each 0.27 mm x 0.27 mm in size, was studied in a microgravity environment provided by NASA's KC-135. Each of the heaters was maintained at a constant temperature by means of electronic feedback circuits, and the heat flux through each individual heater was measured at a high sampling rate. Space and time resolved heat flux maps were obtained and correlated with video pictures of boiling on the surface recorded from below. The time resolved heat flux data was then conditionally sampled according to whether or not boiling occurred on the surface and an average heat flux during boiling was obtained. Array averaged heat fluxes in microgravity were slightly larger than in Earth gravity for wall superheats up to about 30 K, but were significantly lower than in Earth gravity at higher superheats. The time-average heat flux conditionally sampled on boiling, however, was independent of the gravity level suggesting that the behavior of small bubbles is not affected by gravity. Heat transfer from the surface occurred primarily through these small bubbles-not much heat transfer was associated with the large bubble that occasionally formed on the surface as a result of coalescence of the small bubbles. (orig.)

  16. Boiling heat transfer of nanofluids--special emphasis on critical heat flux.

    Science.gov (United States)

    Kim, Sung Joong; Kim, Hyungdae

    2013-11-01

    As innovative nanotechnology-based heat-transfer media, nanofluids have evoked considerable interest among researchers owing to their improved thermal properties as well as their extendable applications to various high-power thermal systems. This paper presents a comprehensive review of recent research developments and patents pertaining to nanofluid boiling heat transfer. Nanofluids definitely offer a wide range of potential improvements in boiling heat-transfer performance. However, experimental data available from different studies are currently beset by numerous contradictions, suggesting that the fundamental mechanisms of nanofluid boiling heat transfer are not yet well understood. Consequently application of these technologies has been limited in some aspects. Only a small number of patents related to nanofluid boiling heat transfer have thus far been reported in the literature. Based on the present review, future technological development and research requirements in this area are outlined in line with technical challenges. To utilize nanofluid boiling heat-transfer technologies for practical applications, more systematic and fundamental studies are required to understand the physical mechanisms involved.

  17. U-Tube steam generator modeling for natural circulation and pot-boiling modes

    International Nuclear Information System (INIS)

    A general steam generator model is developed to account for various modes such as natural circulation and pot-boiling modes which can be occurred during transients. The present model can describe not only the swell and shrink phenomena, occurred by any small change in steam flowrates and feedwater flowrates but also natural circulation, pot-boiling, and tube uncovery modes which occur in sequence during the loss of feedwater transient if auxiliary feedwater is not followed by. The void fraction concept is used instead of the quality concept to simulate counter-current flow which takes place in the pot-boiling mode after the natural circulation mode stops. The present model is based on a one-dimensional three-region model to realistically describe the swell and shrink phenomena; downcomer, tube bundle, and steam dome regions. Both of the downcomer water level and two-phase level can be predicted during the pot-boiling mode. To verify the present model in the pot-boiling mode, a simple experiment is done and simulated by the present code. It is shown that the simulated results are relatively in good agreement with the experimental data

  18. Experimental study on forced convection boiling heat transfer on molten alloy

    International Nuclear Information System (INIS)

    In order to clarify the characteristics of forced convection boiling heat transfer on molten metal, basic experiments have been carried out with subcooled water flowing on molten Wood's alloy pool surface. In these experiments, water flows horizontally in a rectangular duct. A cavity filled with Wood's alloy is present in a portion of the bottom of the duct. Wood's alloy is heated by a copper conductor at the bottom of the cavity. The experiments have been carried out with various velocities and subcoolings of water, and temperature of Wood's alloy. Boiling curves on the molten alloy surface were obtained and compared with that on a solid heat transfer surface. It is observed that the boiling curve on molten alloy is in a lower superheat region than the boiling curve on a solid surface. This indicates that the heat transfer performance of forced convection boiling on molten alloy is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy

  19. Pool boiling of nanofluids on rough and porous coated tubes: experimental and correlation

    Science.gov (United States)

    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.

  20. Pool boiling of nanofluids on rough and porous coated tubes: experimental and correlation

    Directory of Open Access Journals (Sweden)

    Cieśliński Janusz T.

    2014-06-01

    Full Text Available 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.

  1. Numerical Modeling of Propellant Boil-Off in a Cryogenic Storage Tank

    Science.gov (United States)

    Majumdar, A. K.; Steadman, T. E.; Maroney, J. L.; Sass, J. P.; Fesmire, J. E.

    2007-01-01

    A numerical model to predict boil-off of stored propellant in large spherical cryogenic tanks has been developed. Accurate prediction of tank boil-off rates for different thermal insulation systems was the goal of this collaboration effort. The Generalized Fluid System Simulation Program, integrating flow analysis and conjugate heat transfer for solving complex fluid system problems, was used to create the model. Calculation of tank boil-off rate requires simultaneous simulation of heat transfer processes among liquid propellant, vapor ullage space, and tank structure. The reference tank for the boil-off model was the 850,000 gallon liquid hydrogen tank at Launch Complex 39B (LC- 39B) at Kennedy Space Center, which is under study for future infrastructure improvements to support the Constellation program. The methodology employed in the numerical model was validated using a sub-scale model and tank. Experimental test data from a 1/15th scale version of the LC-39B tank using both liquid hydrogen and liquid nitrogen were used to anchor the analytical predictions of the sub-scale model. Favorable correlations between sub-scale model and experimental test data have provided confidence in full-scale tank boil-off predictions. These methods are now being used in the preliminary design for other cases including future launch vehicles

  2. A study on the correlations development for film boiling heat transfer on spheres

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  3. Measurement of boiling heat transfer coefficient in liquid nitrogen bath by inverse heat conduction method

    Institute of Scientific and Technical Information of China (English)

    Tao JIN; Jian-ping HONG; Hao ZHENG; Ke TANG; Zhi-hua GAN

    2009-01-01

    Inverse heat conduction method (IHCM)is one of the most effective approaches to obtaining the boiling heat transfer coefficient from measured results.This paper focuses on its application in cryogenic boiling heat transfer.Experiments were conducted on the heattransfer of a stainless steel block in a liquid nitrogen bath.with the assumption of a ID conduction condition to realize fast acquisition of the temperature of the test points inside the block.With the inverse-heat conduction theory and the explicit finite difference model,a solving program was developed to calculate the heat flux and the boiling heat transfer coefficient of a stainless steel block in liquid nitrogen bath based on the temperature acquisition data.Considering the oscillating data and some unsmooth transition points in the inverse-heat-conduction calculation result of the heat-transfer coefficient,a two-step data-fitting procedure was proposed to obtain the expression for the boiling heat transfer coefficients.The coefficient was then verified for accuracy by a comparison between the simulation results using this expression and the verifying experimental results of a stainless steel block.The maximum error with a revised segment fitting iS around 6%.which verifies the feasibility of using IHCM to measure the boiling heat transfer coefficient in liquid nitrogen bath.

  4. Experimental and theoretical study of pool boiling heat transfer to amine solutions

    Directory of Open Access Journals (Sweden)

    S. M. Peyghambarzadeh

    2009-03-01

    Full Text Available In this investigation, a large number of experiments have been performed to measure the nucleate boiling heat transfer coefficients of water/diethanolamine (DEA and water/monoethanolamine (MEA binary solutions. Based on these experimental data, effects of physical properties such as surface tension and viscosity of mixtures on nucleate boiling heat transfer coefficients and also on bubble dynamics have been discussed. Meanwhile, some photographs have also been selected which illustrate the behaviours of bubbles near the heat transfer surface. In this article, a new correlation has been developed on the basis of the correlation of Stephan and Körner which is known as a successful correlation for the prediction of nucleate boiling heat transfer coefficients of mixtures. Comparison of the prediction of this new correlation with experimental data indicates that this modification can significantly improve the performance of the Stephan and Körner correlation.

  5. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  6. Forced convection flow boiling and two-phase flow phenomena in a microchannel

    Science.gov (United States)

    Na, Yun Whan

    2008-07-01

    The present study was performed to numerically analyze the evaporation phenomena through the liquid-vapor interface and to investigate bubble dynamics and heat transfer behavior during forced convective flow boiling in a microchannel. Flow instabilities of two-phase flow boiling in a microchannel were studied as well. The main objective of this research is to investigate the fundamental mechanisms of two-phase flow boiling in a microchannel and provide predictive tools to design thermal management systems, for example, microchannel heat sinks. The numerical results obtained from this study were qualitatively and quantitatively compared with experimental results in the open literature. Physical and mathematical models, accounting for evaporating phenomena through the liquid-vapor interface in a microchannel at constant heat flux and constant wall temperature, have been developed, respectively. The heat transfer mechanism is affected by the dominant heat conduction through the thin liquid film and vaporization at the liquid-vapor interface. The thickness of the liquid film and the pressure of the liquid and vapor phases were simultaneously solved by the governing differential equations. The developed semi-analytical evaporation model that takes into account of the interfacial phenomena and surface tension effects was used to obtain solutions numerically using the fourth-order Runge-Kutta method. The effects of heat flux 19 and wall temperature on the liquid film were evaluated. The obtained pressure drops in a microchannel were qualitatively consistent with the experimental results of Qu and Mudawar (2004). Forced convective flow boiling in a single microchannel with different channel heights was studied through a numerical simulation to investigate bubble dynamics, flow patterns, and heat transfer. The momentum and energy equations were solved using the finite volume method while the liquid-vapor interface of a bubble is captured using the VOF (Volume of Fluid

  7. High-Speed Visualization of Bubble Behaviors for Pool Boiling of R-141b

    Institute of Scientific and Technical Information of China (English)

    Yanhua DIAO; Yaohua ZHAO; Qiuliang WANG

    2006-01-01

    A visualization study on the behavior of bubbles has been carried out for pool boiling of R141b on a horizontal transparent heater at pressure 0.1 MPa. The behaviors of bubbles were recorded by a high-speed camera placed beneath the heater surface. The departure diameter, departure time of bubbles and nucleation site density at different heat flux were obtained. The visualization results show that bubble departure diameter and departure time decrease, while the nucleation site density increases as the heat flux increases. It is also observed that there is no liquid recruited into the microlayer in the experiment. Based on the experimental results, boiling curve for R141b was predicted by using the dynamic microlayer model. As a result, the agreement between the predictive result based on the dynamic microlayer model and the experiment data for boiling curve of R141b is good at high heat flux.

  8. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  9. Determination of local boiling in light water reactors by correlation of the neutron noise

    International Nuclear Information System (INIS)

    The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author)

  10. The effect of a gas dissolved in a coolant on boiling and burnout heat transfer

    International Nuclear Information System (INIS)

    Experiments were conducted to determine the effect of nitrogen dissolved in water on boiling and burnout heat transfer in a cylindrical pipe. Gas liberation from a liquid was found to enhance heat transfer. In the region of subcooled boiling a reduction of the critical heat flux up to about 20% was observed when in the region of burnout development the coolant was degassed at rather a high rate (high pressure and mass flow rates). As the subcooling becomes smaller the difference between q/sub cr/ for a degassed and a gas-saturated boiling coolant decreases. On the attainment of a thermally equilibrium two-phase flow the values of q/sub cr/ practically coincide for both cases

  11. Study of density wave phenomena in boiling and condensing two-phase flow systems

    International Nuclear Information System (INIS)

    In this work density wave oscillations are studied. This phenomenon have been widely studied for boiling systems with sub-cooled inlet condition in the past. The main purpose of this work is to characterize the stability region of boiling and condensing systems for sub-cooled and saturated inlet conditions. Stability maps, based on the sub-cooling and the phase-change numbers, are constructed. The limits of the unstable regions are identified and characterized. Finally some numerical simulations are presented in order to describe the nature of the involved phenomena. A high-order numerical solver, based on a homogenous two-phase model for a single boiling channel is implemented. (author)

  12. Quantitative measurement of void fraction in a forced convective flow boiling by using neutron radiography

    International Nuclear Information System (INIS)

    The void fraction in a forced convective flow boiling is very important information for understanding the characteristics of the boiling two-phase flow. Consequently, many experimental investigations have been carried out to obtain the local void fraction so far, but the detail data among the whole of the test-section has not been enough. Especially, the data under subcooled condition are quite limited. In this study, the void fraction distribution in a forced convective boiling was quantitatively measured by using the thermal neutron radiography. These results were compared with several existing void fraction correlations. Although these correlations show a good agreements with experimental results under low heat flux condition, there is no suitable correlation to estimate the void fraction under non-thermal equilibrium condition. (author)

  13. Verification of the IVA4 film boiling model with the data base of Liu and Theofanous

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1998-01-01

    Part 1 of this work presents a closed analytical solution for mixed-convection film boiling on vertical walls. Heat transfer coefficients predicted by the proposed model and experimental data obtained at the Royal Institute of Technology in Sweden by Okkonen et al are compared. All data predicted are inside the {+-}10% error band, with mean averaged error being below 4% using the slightly modified analytical solution. The solution obtained is recommended for practical applications. The method presented here is used in Part 2 as a guideline for developing model for film boiling on spheres. The new semi-empirical film boiling model for spheres used in IVA4 computer code is compared with the experimental data base obtained by Liu and Theofanous. The data are predicted within {+-}30% error band. (author)

  14. Characteristics of liquid and boiling sodium flows in heating pin bundles

    International Nuclear Information System (INIS)

    This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged

  15. Dynamics of Discrete Bubble in Nucleate Pool Boiling on Thin Wires in Micro-gravity

    Institute of Scientific and Technical Information of China (English)

    Shixin WAN; Jianfu ZHAO; Gang LIU

    2009-01-01

    A space experiment on bubble behavior and heat transfer in subcooled pool boiling phenomenon has been per-formed utilizing the temperature-controlled pool boiling (TCPB) device both in normal gravity in the laboratory and in microgravity aboard the 22no Chinese recoverable satellite. The fluid is degassed Rl13 at 0.1 Mpa and subcooled by 26℃ nominally. A thin platinum wire of 60 μm in diameter and 30 mm in length is simultaneously used as heater and thermometer. Only the dynamics of the vapor bubbles, particularly the lateral motion and the departure of discrete vapor bubbles in nucleate pool boiling are reported and analyzed in the present paper. It's found that these distinct behaviors can be explained by the Marangoni convection in the liquid surrounding vapor bubbles. The origin of the Marangoni effect is also discussed.

  16. Heat Transfer From Electrically Heated Nichrome Wires to Boiling Water at Different Pressures

    Directory of Open Access Journals (Sweden)

    Devi Dayal

    1968-01-01

    Full Text Available Boiling curves for nucleate and film boiling have been drawn for nichrome of three sizes in distilled and degasified water at saturation temperatures under five different sub-atmospheric vapour pressure. It has been observed that (i for the same Q/A (heat transfer, Delta Theta (excess of wire temperature over saturation point of water decreases with pressure in both nucleate and film boiling ranges, (ii Both Q/A max. and Delta Theta/SubC show a rapid decrease with pressure but these variations become more gradual at higher pressures, and (iii Q/A max. and Delta Theta/SubC increase with wire size at all pressures; increase in Delta Theta/SubC however, becomes less conspicuous at higher pressures approaching one atmosphere.

  17. An improved mechanistic critical heat flux model for subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

  18. Modeling and measurement of boiling point elevation during water vaporization from aqueous urea for SCR applications

    Energy Technology Data Exchange (ETDEWEB)

    Dan, Ho Jin; Lee, Joon Sik [Seoul National University, Seoul (Korea, Republic of)

    2016-03-15

    Understanding of water vaporization is the first step to anticipate the conversion process of urea into ammonia in the exhaust stream. As aqueous urea is a mixture and the urea in the mixture acts as a non-volatile solute, its colligative properties should be considered during water vaporization. The elevation of boiling point for urea water solution is measured with respect to urea mole fraction. With the boiling-point elevation relation, a model for water vaporization is proposed underlining the correction of the heat of vaporization of water in the urea water mixture due to the enthalpy of urea dissolution in water. The model is verified by the experiments of water vaporization as well. Finally, the water vaporization model is applied to the water vaporization of aqueous urea droplets. It is shown that urea decomposition can begin before water evaporation finishes due to the boiling-point elevation.

  19. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    International Nuclear Information System (INIS)

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

  20. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    Energy Technology Data Exchange (ETDEWEB)

    Aksan, S.N.; Stierli, F.; Analytis, G.T. [Paul Scherrer Inst. (PSI), Villigen (Switzerland). Lab. for Thermal-Hydraulics

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended.

  1. EXAMINATION OF THE CERTAIN CHEMICAL CHARACTERISTICS OF DIFFERENT TYPES OF BOILED SAUSAGES PRODUCED IN SERBIA

    Directory of Open Access Journals (Sweden)

    Vladimir Kurćubić

    2012-12-01

    Full Text Available The objectives of our study were to examine certain chemical quality parameters in samples of various subgroups of boiled sausages and compare obtained values with reference values (Rule book on the quality of minced meat, semi-finished meat and meat products, Official Gazette of RS, no. 31/2012. We used two reference chemical methods: determination of nitrogen and phosphorus content (SRPS ISO 937:1992, SRPS ISO 13730:1999. For determination of hydroxyproline we used M 050 “in house” modified method. We concluded that the total number of tested samples of Fine grounded boiled sausages (n=170, 47 (27.65% does not match the requirements for the quality provided the above mentioned Rule book, of which 21 samples because lower total protein content (TP and 33 samples because higher relative protein content of connective tissue (RPC. The most common chemical quality failure of examined Coarsely grounded boiled sausages (n=94 is a higher percent of RPC than allowed (18 - 25 %. We consider in samples of same subgroup lower content of TP in 8 samples (11.11 %. From a total of 8 samples tested Boiled sausages with meat peaces (Šunkarica, all tested samples revealed lower content of TP than allowed (100 % and 2 samples (25 % higher relative protein content of connective tissue (RPC. The most common deficiency that reduces the chemical quality of the boiled sausages is higher level of RPC, demonstrated in 53 (19.48 % of the total (n=272 samples tested. Something minor drawback is the lower percent of TP determined in 37 samples (13.60 %. Total P2O5 content was compatibile with the values permitted by the Rule book in all of 272 examined samples, indicating a strict adherence to regulations when using phosphate as one of the technologically most important additives in the production of boiled sausages.

  2. Bench-scale screening tests for a boiling sodium-potassium alloy solar receiver

    Science.gov (United States)

    Moreno, J. B.; Moss, T. A.

    1993-06-01

    Bench-scale tests were carried out in support of the design of a second-generation 75-kW(sub t) reflux pool-boiler solar receiver. The receiver will be made from Haynes Alloy 230 and will contain the sodium-potassium alloy NaK-78. The bench-scale tests used quartz lamp heated boilers to screen candidate boiling stabilization materials and methods at temperatures up to 750 degree C. Candidates that provided stable boiling were tested for hot-restart behavior. Poor stability was obtained with single 1/4-inch diameter patches of powdered metal hot press sintered onto the wetted side of the heat-input area. Laser-drilled and electric discharge machined cavities in the heated surface also performed poorly. Small additions of xenon, and heated-surface tilt out of the vertical, dramatically improved poor boiling stability; additions of helium or oxygen did not. The most stable boiling was obtained when the entire heat-input area was covered by a powdered-metal coating. The effect of heated-area size was assessed for one coating: at low incident fluxes, when even this coating performed poorly, increasing the heated-area size markedly improved boiling stability. Good hot-restart behavior was not observed with any candidate, although results were significantly better with added xenon in a boiler shortened from 3 to 2 feet. In addition to the screening tests, flash-radiography imaging of metal-vapor bubbles during boiling was attempted. Contrary to the Cole-Rohsenow correlation, these bubble-size estimates did not vary with pressure; instead they were constant, consistent with the only other alkali metal measurements, but about 1/2 their size.

  3. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  4. Specific interface area and self-stirring in a two-liquid system experiencing intense interfacial boiling below the bulk boiling temperatures of both components

    CERN Document Server

    Goldobin, Denis S

    2016-01-01

    We present an approach to theoretical assessment of the mean specific interface area $(\\delta{S}/\\delta{V})$ for a well-stirred system of two immiscible liquids experiencing interfacial boiling. The assessment is based on the balance of transformations of mechanical energy and the laws of the momentum and heat transfer in the turbulent boundary layer. The theory yields relations between the specific interface area and the characteristics of the system state. In particular, this allows us to derive the equations of self-cooling dynamics of the system in the absence of external heat supply. The results provide possibility for constructing a self-contained mathematical description of the process of interfacial boiling. In this study, we assume the volume fractions of two components to be similar as well as the values of their kinematic viscosity and molecular heat diffusivity.

  5. Simple Alcohols with the Lowest Normal Boiling Point Using Topological Indices

    OpenAIRE

    Goubko, Mikhail; Miloserdov, Oleg

    2015-01-01

    We find simple saturated alcohols with the given number of carbon atoms and the minimal normal boiling point. The boiling point is predicted with a weighted sum of the generalized first Zagreb index, the second Zagreb index, the Wiener index for vertex-weighted graphs, and a simple index caring for the degree of a carbon atom being incident to the hydroxyl group. To find extremal alcohol molecules we characterize chemical trees of order $n$, which minimize the sum of the second Zagreb index a...

  6. Analysis and Measurement of Bubble Dynamics and Associated Flow Field in Subcooled Nucleate Boiling Flows

    Energy Technology Data Exchange (ETDEWEB)

    Barclay G. Jones

    2008-10-01

    In recent years, subooled nucleate boiling (SNB) has attrcted expanding research interest owing to the emergence of axial offset anomaly (AOA) or crud-induced power shigt (CIPS) in many operating US PWRs, which is an unexpected deviation in the core axial power distribution from the predicted power curves. Research indicates that the formation of the crud, which directly leads to AOA phenomena, results from the presence of the subcooled nucleate boiling, and is especially realted to bubble motion occurring in the core region.

  7. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  8. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  9. Modelling and validation of turbulent boiling flow in a rectangular channel

    International Nuclear Information System (INIS)

    The boiling flow experiments in a rectangular vertical channel, performed at the Texas A&M University, were used to validate the prediction capabilities of the NEPTUNECFD code. The refrigerant HFE-301 was used as a working fluid. Liquid velocity and turbulent kinetic energy profiles close to the heated wall were compared. NEPTUNECFD simulations successfully predict the main experimental tendencies associated with the heat flux and Reynolds number variation. However, a disagreement in velocity and turbulence magnitude can be observed in the boiling region at lower Reynolds numbers. The size of experimental bubble diameter in the vicinity of the heated wall is assessed by the means of non-dimensional analysis. (author)

  10. Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

    International Nuclear Information System (INIS)

    In this work, we report on the development of a new computer code named TRACE for predicting the excursion characteristics of criticality excursions involving fissile solutions. TRACE employs point neutronics coupled with simple thermal-hydraulics. The temperature, the radiolytic gas effects, and the boiling phenomena are estimated using the transient heat conduction equation, a lumped-parameter energy model, and a simple boiling model, respectively. To evaluate the model, we compared our results with the results of CRAC experiments. The agreement in these comparisons is quite satisfactory. (author)

  11. Application of fractal characteristic quantities of pressure fluctuation in subcooled boiling regime recognition

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The dynamical behavior of the subcoole d boiling two-phase system ws introduced and discussed. With the introduction of fractal concept, an analysis of the fractal feature of pressure wave signals fiom nonlinear dynamics point of view. was carried out. Meanwhile, the pseudo phase diagrans of typical time series of sound pressure were given. Finally, through dynamic clustering and on the basis of calculating correlation dimension and Hurst exponent of pressure wave time series on different subcooling conditions, the recognition of developing regime of the two-phase system was delivered, which might provide a promising approach of recognition and diagnosis of a boiling system.

  12. Reexamination of Correlations for Nucleate Site Distribution on Boiling Surface by Fractal Theory

    Institute of Scientific and Technical Information of China (English)

    YangChunxin

    1997-01-01

    Nucleate site distribution plays an essential role in nucleate boiling process.In this paper,it is pointed out that the size and spatial distributioin density of nucleate sites presented on real boiling surface can be described by the normalized fractal distribution function,and the physical meaning of parameters involved in some experimental correlations proposed by early investigations are identified according to fractal distribution function.It is further suggested that the surface micro geometry characteristics such as the shape of cavities should be described and analyzed qualitatively by using fractal theory.

  13. Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

    Energy Technology Data Exchange (ETDEWEB)

    Basoglu, Benan; Yamamoto, Toshihiro; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In this work, we report on the development of a new computer code named TRACE for predicting the excursion characteristics of criticality excursions involving fissile solutions. TRACE employs point neutronics coupled with simple thermal-hydraulics. The temperature, the radiolytic gas effects, and the boiling phenomena are estimated using the transient heat conduction equation, a lumped-parameter energy model, and a simple boiling model, respectively. To evaluate the model, we compared our results with the results of CRAC experiments. The agreement in these comparisons is quite satisfactory. (author)

  14. Estimation of normal boiling points of trialkyl phosphates using retention indices by gas chromatography

    International Nuclear Information System (INIS)

    Retention indices of several homologous trialkyl phosphates have been determined by gas chromatography on different polar stationary phases namely, Apiezon L, SE-30 and XE-60. Normal boiling points of these trialkyl phosphates have been evaluated and compared with available literature values. Topological indices such as Xu index, atom type index and steric effect index are derived for these phosphates and have been correlated with the normal boiling points using multiple regression analysis. The influences of alkyl chain length, relative position of alkyl branching and steric factors on retention index are investigated and also the effect of polarity of the stationary phase on retention indices is discussed.

  15. Assessment of Nucleation Site Density Models for CFD Simulations of Subcooled Flow Boiling

    International Nuclear Information System (INIS)

    The framework of a CFD simulation of subcooled flow boiling basically includes a block of wall boiling models communicating with governing equations of a two-phase flow via parameters like temperature, rate of phasic change, etc. In the block of wall boiling models, a heat flux partitioning model, which describes how the heat is taken away from a heated surface, is combined with models quantifying boiling parameters, i.e. nucleation site density, and bubble departure diameter and frequency. It is realized that the nucleation site density is an important parameter for predicting the subcooled flow boiling. The number of nucleation sites per unit area decides the influence region of each heat transfer mechanism. The variation of the nucleation site density will mutually change the dynamics of vapor bubbles formed at these sites. In addition, the nucleation site density is needed as one initial and boundary condition to solve the interfacial area transport equation. A lot of effort has been devoted to mathematically formulate the nucleation site density. As a consequence, numerous correlations of the nucleation site density are available in the literature. These correlations are commonly quite different in their mathematical form as well as application range. Some correlations of the nucleation site density have been applied successfully to CFD simulations of several specific subcooled boiling flows, but in combination with different correlations of the bubble departure diameter and frequency. In addition, the values of the nucleation site density, and bubble departure diameter and frequency obtained from simulations for a same problem are relatively different, depending on which models are used, even when global characteristics, e.g., void fraction and mean bubble diameter, agree well with experimental values. It is realized that having a good CFD simulations of the subcooled flow boiling requires a detailed validations of all the models used. Owing to the importance

  16. Decay heat removal under boiling condition in a pin-bundle geometry

    International Nuclear Information System (INIS)

    Decay heat removal capability under boiling condition was investigated using an electrically heated 37-pin bundle test section. The flow was driven by natural circulation force of the out-of-pile sodium loop SIENA in O-arai Engineering Center, PNC. As the heater power was increased, the two-phase flow regime changed from bubbly flow to slug flow and then to annular or annular mist flow. In 15 runs, dry-out was not observed in the average exit quality region of less than 0.5. The results indicated the existance of a large ''boiling window'' for low flow rate and low power conditions. (author)

  17. Parameter sensitivity study of boiling and two-phase flow models in computational thermal hydraulics

    International Nuclear Information System (INIS)

    This work presents a sensitivity study of boiling and two phase flow models for thermal hydraulics simulations in nuclear reactors. The study quantifies sources of uncertainty and error in these simulations by computing global sensitivities of figures of merit, or outputs, to model parameters, inputs, and mesh resolution. Results are obtained for the DEBORA benchmark problem of boiling in a channel driven by a heated wall section. Scalar outputs of interest are average wall temperature, integrated cross-sectional void fraction, and pressure drop in the channel. Sensitivities are computed with respect to both individual heat fluxes and to the parameters in the models for these heat fluxes. (author)

  18. Effect of different technological processes on decontamination of meat contaminated with radiocesium Pt. 1. Use of boiling

    International Nuclear Information System (INIS)

    The decontamination effect of boiling of pork and mutton contaminated with Cs 137 was investigated. The results of the study indicate that the decontamination effectiveness of boiling pork and mutton for 2 hours to about 80 and 76 per cent respectively. (author). 10 refs, 2 tabs

  19. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Science.gov (United States)

    2010-05-13

    ... HUMAN SERVICES Food and Drug Administration Guidance for Industry: Use of Water by Food Manufacturers in... entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water... ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory.''...

  20. Effect of soaking, boiling, and steaming on total phenolic contentand antioxidant activities of cool season food legumes.

    Science.gov (United States)

    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 (pyellow 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. PMID:26050159

  1. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O2; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  2. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    General design criteria for nuclear power plants in every OECD country require that the reactor core and associated coolant, control, and protection systems be designed so that power oscillations which can result in conditions exceeding acceptable fuel design limits are not possible, or they can be reliably and readily detected and suppressed. In practice, this means that reactor cores should be stable with regard to perturbations from their normal operating state, so that expected variations to the operating parameters do not induce undamped power oscillations. These power oscillations can take a variety of forms, from very local power peaks which can cause no damage, or only slight damage to only a few fuel rods, to large core-wide oscillations where entire segments of the core can become neutronically uncoupled, with wide power swings. Ever since the fast boiling water reactors began operating, over 30 years ago, it has been recognized that their operation under certain conditions of power and flow could cause power and flow oscillations. Considerable research was performed at that time to better understand the principal operating parameters which contribute to the initiation of these oscillations, and guidelines were developed to avoid plant operation under the conditions which were the most unstable. Experiments in the the first Special Power Excursion Reactor Test (SPERT-1) program produced spontaneous power oscillations, and investigations in an out-of-pile loop were necessary to demonstrate that the immediate cause of the oscillations was a power-to-reactivity feedback. Further investigations indicated that the instabilities were limited to certain areas on the operating map. These regions could not be absolutely defined, but there was sufficient understanding of them that they could be generally avoided, with only minor examples of instability events. More recently, though, several reactor events, and especially one that occurred at the La Salle Nuclear

  3. Subcooled flow boiling heat transfer of ethanol aqueous solutions in vertical annulus space

    Directory of Open Access Journals (Sweden)

    Sarafraz M.M.

    2012-01-01

    Full Text Available The subcooled flow boiling heat-transfer characteristics of water and ethanol solutions in a vertical annulus have been investigated up to heat flux 132kW/m2. The variations in the effects of heat flux and fluid velocity, and concentration of ethanol on the observed heat-transfer coefficients over a range of ethanol concentrations implied an enhanced contribution of nucleate boiling heat transfer in flow boiling, where both forced convection and nucleate boiling heat transfer occurred. Increasing the ethanol concentration led to a significant deterioration in the observed heat-transfer coefficient because of a mixture effect, that resulted in a local rise in the saturation temperature of ethanol/water solution at the vapor-liquid interface. The reduction in the heat-transfer coefficient with increasing ethanol concentration is also attributed to changes in the fluid properties (for example, viscosity and heat capacity of tested solutions with different ethanol content. The experimental data were compared with some well-established existing correlations. Results of comparisons indicate existing correlations are unable to obtain the acceptable values. Therefore a modified correlation based on Gnielinski correlation has been proposed that predicts the heat transfer coefficient for ethanol/water solution with uncertainty about 8% that is the least in comparison to other well-known existing correlations.

  4. Bubble Departure from Metal-Graphite Composite Surfaces and Its Effects on Pool Boiling Heat Transfer

    Science.gov (United States)

    Chao, David F.; Sankovic, John M.; Motil, Brian J.; Yang, W-J.; Zhang, Nengli

    2010-01-01

    The formation and growth processes of a bubble in the vicinity of graphite micro-fiber tips on metal-graphite composite boiling surfaces and their effects on boiling behavior are investigated. It is discovered that a large number of micro bubbles are formed first at the micro scratches and cavities on the metal matrix in pool boiling. By virtue of the non-wetting property of graphite, once the growing micro bubbles touch the graphite tips, the micro bubbles are sucked by the tips and merged into larger micro bubbles sitting on the end of the tips. The micro bubbles grow rapidly and coalesce to form macro bubbles, each spanning several tips. The necking process of a detaching macro bubble is analyzed. It is revealed that a liquid jet is produced by sudden break-off of the bubble throat. The composite surfaces not only have higher temperatures in micro- and macrolayers but also make higher frequency of the bubble departure, which increase the average heat fluxes in both the bubble growth stage and in the bubble departure period. Based on these analyses, the enhancement mechanism of pool boiling heat transfer on composite surfaces is clearly revealed.

  5. Explaining Melting and Evaporation below Boiling Point. Can Software Help with Particle Ideas?

    Science.gov (United States)

    Papageorgiou, George; Johnson, Philip; Fotiades, Fotis

    2008-01-01

    This paper reports the findings of a study exploring the use of a software package to help pupils understand particulate explanations for melting and evaporation below boiling point. Two matched classes in a primary school in Greece (ages 11-12, n = 16 and 19) were involved in a short intervention of six one hour lessons. Covering the same…

  6. Boiling wax burn in mid-autumn festival in Hong Kong.

    Science.gov (United States)

    Chan, E S; Chan, E C; Ho, W S; King, W W

    1997-01-01

    An unusual cause of burn, contact with boiling wax by children and adolescents during the annual mid-autumn festival in Hong Kong is presented. 57 patients who suffered from hot wax burn over the period 1986-1996 were admitted to the Burns Unit of the Prince of Wales Hospital. This special burn should be preventable by public education.

  7. What Is the Boiling Point and Heat of Vaporization of Sulfuric Acid?

    Science.gov (United States)

    Myers, R. Thomas

    1983-01-01

    Discusses the values presented in various handbooks for the boiling point and heat of vaporization of sulfuric acid, noting discrepencies. Analyzes various approaches to data presentation, discussing the data on sulfuric acid in light of the Trouton constant. Points out the need for a more critical use of tables. (JM)

  8. The Analysis of Existence and Influence of the Boiling Point at Vacuum Drying

    OpenAIRE

    NEGRĂU Petru Mircea

    2011-01-01

    Accurate determination of the existence,change and influence of the boiling point of waterinside the material who will be dried is a primarydomain. In this article will be displayed and analyzedrelatively the results of the in-house experiments onvacuum drying wood using radio-frequency waves.

  9. Kinetics-based phase change approach for VOF method applied to boiling flow

    Science.gov (United States)

    Cifani, Paolo; Geurts, Bernard; Kuerten, Hans

    2014-11-01

    Direct numerical simulations of boiling flows are performed to better understand the interaction of boiling phenomena with turbulence. The multiphase flow is simulated by solving a single set of equations for the whole flow field according to the one-fluid formulation, using a VOF interface capturing method. Interface terms, related to surface tension, interphase mass transfer and latent heat, are added at the phase boundary. The mass transfer rate across the interface is derived from kinetic theory and subsequently coupled with the continuum representation of the flow field. The numerical model was implemented in OpenFOAM and validated against 3 cases: evaporation of a spherical uniformly heated droplet, growth of a spherical bubble in a superheated liquid and two dimensional film boiling. The computational model will be used to investigate the change in turbulence intensity in a fully developed channel flow due to interaction with boiling heat and mass transfer. In particular, we will focus on the influence of the vapor bubble volume fraction on enhancing heat and mass transfer. Furthermore, we will investigate kinetic energy spectra in order to identify the dynamics associated with the wakes of vapor bubbles. Department of Applied Mathematics, 7500 AE Enschede, NL.

  10. Boiling peanut Ara h 1 results in the formation of aggregates with reduced allergenicity

    NARCIS (Netherlands)

    F. Blanc; Y.M. Vissers; K. Adel-Patient; N.M. Rigby; A.R. Mackie; A.P. Gunning; N.K. Wellner; P.S. Skov; L. Przybylski-Nicaise; B. Ballmer-Weber; L. Zuidmeer-Jongejan; Z. Szepfalusi; J. Ruinemans-Koerts; A.P.H. Jansen; H. Bernard; J.M. Wal; H.F.J. Savelkoul; H.J. Wichers; E.N.C. Mills

    2011-01-01

    Scope: Roasting rather than boiling and Maillard modifications may modulate peanut allergenicity. We investigated how these factors affect the allergenic properties of a major peanut allergen, Ara h 1. Methods and results: Ara h 1 was purified from either raw (N-Ara h 1) or roasted (R-Ara h 1) peanu

  11. Investigation into flow boiling heat transfer in a minichannel with enhanced heating surface

    Directory of Open Access Journals (Sweden)

    Piasecka Magdalena

    2012-04-01

    Full Text Available The paper presents results of flow boiling in a minichannel of 1.0 mm depth. The heating element for the working fluid (FC-72 that flows along the minichannel is a single-sided enhanced alloy foil made from Haynes-230. Microrecesses were formed on the selected area of the heating foil by laser technology. The observations of the flow structure were carried out through a piece of glass. Simultaneously, owing to the liquid crystal layer placed on the opposite side of the enhanced foil surface, it was possible to measure temperature distribution on the heating wall through another piece of glass. The experimental research has been focused on the transition from single phase forced convection to nucleate boiling, i.e. the zone of boiling incipience and further development of boiling. The objective of the paper is determining of the void fraction for some cross-sections of selected images for increasing heat fluxes supplied to the heating surface. The flow structure photos were processed in Corel graphics software and binarized. The analysis of phase volumes was developed in Techystem Globe software.

  12. Investigation into flow boiling heat transfer in a minichannel with enhanced heating surface

    Science.gov (United States)

    Piasecka, Magdalena

    2012-04-01

    The paper presents results of flow boiling in a minichannel of 1.0 mm depth. The heating element for the working fluid (FC-72) that flows along the minichannel is a single-sided enhanced alloy foil made from Haynes-230. Microrecesses were formed on the selected area of the heating foil by laser technology. The observations of the flow structure were carried out through a piece of glass. Simultaneously, owing to the liquid crystal layer placed on the opposite side of the enhanced foil surface, it was possible to measure temperature distribution on the heating wall through another piece of glass. The experimental research has been focused on the transition from single phase forced convection to nucleate boiling, i.e. the zone of boiling incipience and further development of boiling. The objective of the paper is determining of the void fraction for some cross-sections of selected images for increasing heat fluxes supplied to the heating surface. The flow structure photos were processed in Corel graphics software and binarized. The analysis of phase volumes was developed in Techystem Globe software.

  13. Occurrence conditions of sustainable minute bubble emission boiling for high heat flux cooling

    International Nuclear Information System (INIS)

    In order to investigate the conditions causing minute bubble emission boiling, critical heat flux experiments were conducted in the pool boiling system with different thermal capacities of heat transfer surfaces and method of heating. Stable minute bubble emission boiling was observed for a 10 mm-thick copper cylinder heated by thermal radiation. The critical heat flux obtained was 6.0 MW/m2. When the heat flux exceeded above approximately 3 MW/m2, a large vapor bubble formed on the heat transfer surface, then was condensed immediately and dispersed into minute bubbles. A 4 mm-thick silicon carbide heat transfer surface burnout at 1.9 MW/m2. When the heat transfer surface has a small thermal capacity such as the metal foil, commonly used in CHF experiments, a temporary loss of heat removal due to large bubble formation will cause a rapid temperature increase and will result in burnout. Minute bubble emission boiling could occur stably when the time constant determined by the heating method and thermal capacity exceeds the time required for a large bubble to condense in the subcooled fluid. (author)

  14. Single-bubble dynamics in pool boiling of one-component fluids

    KAUST Repository

    Xu, Xinpeng

    2014-06-04

    We numerically investigate the pool boiling of one-component fluids with a focus on the effects of surface wettability on the single-bubble dynamics. We employed the dynamic van der Waals theory [Phys. Rev. E 75, 036304 (2007)], a diffuse-interface model for liquid-vapor flows involving liquid-vapor transition in nonuniform temperature fields. We first perform simulations for bubbles on homogeneous surfaces. We find that an increase in either the contact angle or the surface superheating can enhance the bubble spreading over the heating surface and increase the bubble departure diameter as well and therefore facilitate the transition into film boiling. We then examine the dynamics of bubbles on patterned surfaces, which incorporate the advantages of both hydrophobic and hydrophilic surfaces. The central hydrophobic region increases the thermodynamic probability of bubble nucleation while the surrounding hydrophilic region hinders the continuous bubble spreading by pinning the contact line at the hydrophobic-hydrophilic intersection. This leads to a small bubble departure diameter and therefore prevents the transition from nucleate boiling into film boiling. With the bubble nucleation probability increased and the bubble departure facilitated, the efficiency of heat transfer on such patterned surfaces is highly enhanced, as observed experimentally [Int. J. Heat Mass Transfer 57, 733 (2013)]. In addition, the stick-slip motion of contact line on patterned surfaces is demonstrated in one-component fluids, with the effect weakened by surface superheating.

  15. Heat transport in boiling turbulent Rayleigh-B\\'{e}nard convection

    CERN Document Server

    Lakkaraju, Rajaram; Oresta, Paolo; Verzicco, Roberto; Lohse, Detlef; Prosperetti, Andrea

    2014-01-01

    Boiling is an extremely effective way to promote heat transfer from a hot surface to a liquid due to several mechanisms many of which are not understood in quantitative detail. An important component of the overall process is that the buoyancy of the bubbles compounds with that of the liquid to give rise to a much enhanced natural convection. In this paper we focus specifically on this enhancement and present a numerical study of the resulting two-phase Rayleigh-B\\'enard convection process. We make no attempt to model other aspects of the boiling process such as bubble nucleation and detachment. We consider a cylindrical cell with a diameter equal to its height. The cell base and top are held at temperatures above and below the boiling point of the liquid, respectively. By keeping the temperature difference constant and changing the liquid pressure we study the effect of the liquid superheat in a Rayleigh number range that, in the absence of boiling, would be between $2\\times10^6$ and $5\\times10^9$. We find a...

  16. Consumers' perception and acceptance of boiled and fermented sausages from strongly boar tainted meat.

    Science.gov (United States)

    Meier-Dinkel, Lisa; Gertheiss, Jan; Schnäckel, Wolfram; Mörlein, Daniel

    2016-08-01

    Characteristic off-flavours may occur in uncastrated male pigs depending on the accumulation of androstenone and skatole. Feasible processing of strongly tainted carcasses is challenging but gains in importance due to the European ban on piglet castration in 2018. This paper investigates consumers' acceptability of two sausage types: (a) emulsion-type (BOILED) and (b) smoked raw-fermented (FERM). Liking (9 point scales) and flavour perception (check-all-that-apply with both, typical and negatively connoted sensory terms) were evaluated by 120 consumers (within-subject design). Proportion of tainted boar meat (0, 50, 100%) affected overall liking of BOILED, F (2, 238)=23.22, P<.001, but not of FERM sausages, F (2, 238)=0.89, P=.414. Consumers described the flavour of BOILED-100 as strong and sweaty. In conclusion, FERM products seem promising for processing of tainted carcasses whereas formulations must be optimized for BOILED in order to eliminate perceptible off-flavours. Boar taint rejection thresholds may be higher for processed than those suggested for unprocessed meat cuts. PMID:27038338

  17. EHD enhancement of pool and in-tube boiling of alternate refrigerants

    Science.gov (United States)

    Ohadi, M. M.; Dessiatoun, S.; Singh, A.; Fanni, M. A.

    1993-08-01

    The electrohydrodynamic (EHD) is an active heat transfer augmentation technique which utilizes the effect of secondary motions generated through the application of an electrostatic potential to a dielectric fluid. Net result is better momentum and heat transfer between the fluid and the heat transfer wall through destabilization of the thermal boundary layer and better mixing of the fluid adjacent to the heat transfer surface. EHD enhancement of refrigerant/refrigerant oil mixtures heat transfer using the electrohydrodynamic (EHD) technique is the subject of a three-year experimental investigation in a project funded by the U.S. Department of Energy, effective June 1, 1993. For the interim period between November 1992 and June 1993 when the DOE funds became available, the Air-Conditioning and Refrigeration Technology Institute (ARTI) provided partial funding for our EHD research program with the aim of accomplishing three major tasks: (1) conduct a comprehensive search of the literature on EHD-enhanced, in-tube and external boiling heat transfer enhancement of alternate refrigerants; (2) design, fabricate, and instrument an in-tube, EHD-enhanced boiling/condensation test rig and perform preliminary testing of the setup; and (3) conduct experiments and document new findings on EHD-enhanced external boiling of alternate refrigerants/refrigerant mixtures in an existing pool boiling test rig apparatus. Description of tasks performed are described and results are discussed.

  18. Transient swelling of liquid level during pool boiling in an emergency condenser

    International Nuclear Information System (INIS)

    The problem of transient swelling in the level of a boiling liquid, encountered in emergency steam condensers of nuclear power plants, reboilers and evaporators has been solved for the first time. This will help determine the space avalable for disengagement of the entrapped liquid droplets in a given sized unit. 6 refs

  19. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition, numerica

  20. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section 73.55... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI...

  1. Hard-Boiled for Hard Times in Leonardo Padura Fuentes's Detective Fiction

    Science.gov (United States)

    Song, H. Rosi

    2009-01-01

    Focusing on Leonardo Padura Fuentes's hard-boiled fiction, this essay traces the origin and evolution of the genre in Cuba. Padura Fuentes has challenged the officially sanctioned socialist "literatura policial" that became popular in the 1970s and 1980s. creating a new model of criticism that is not afraid to confront the island's socio-economic…

  2. Phenolic compouds with antiradical activity from the cork boiling wastewater anaerobic digestion

    OpenAIRE

    Marques, Isabel Paula Ramos; Gil, Luís; La Cara, F

    2013-01-01

    This work aims to develop a procedure that explores the different types of valorization that can be obtained by integrating a biological process, such as the anaerobic digestion, to promote the bioconversion of the industrial cork effluents (cork boiling wastewater, CBW).

  3. Transient measurement of temperature oscillation during noisy film boiling in superfluid helium II

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Peng

    2001-01-01

    [1]Kobayashi, H.?Yasukochi, K., Maximum and minimum heat flux and temperature fluctuation in film-boiling states in superfluid helium, Adv. Cryog. Eng., 1980, 25: 372.[2]Kobayashi, H.?Yasukochi, K., A sample configuration effect on the heat transfer from metal surfaces to pressurized He II, Proc. ICEC, 1980, 8: 217.[3]Schwerdtner, M. V., Stamm, G., Tosi, A. N. et al. The boiling-up process in He II. Optical measurements and visualization, Cryogenics, 1992, 32: 775.[4]Schwerdtner, M. V., Poppes, W., Schmidt, D. W., Distortion of temperature signals in He II due to probe geometry, and a new improved probe, Cryogenics, 1989, 29: 132.[5]Shimazaki, T., Murakami, M.?Iida, T., Second sound wave heat transfer, thermal boundary layer formation and boiling: highly transient heat transport phenomena in He II, Cryogenics, 1995, 35: 645.[6]Zhang, P., Study of physical mechanism of film boiling in He II, Doctoral dissertation, Shanghai Jiaotong University, China, 1998.[7]Arp, V., State equation of liquid helium-4 from 0.8 to 2.5K, J. Low Temp. Phys., 1990, 79: 93.[8]Zhang, P., Kimura, S., Murakami, M. et al., Non-planar and non-linear second sound wave in He II, Chinese Physics Letters, 2000, 17: 43.

  4. Flow boiling critical heat flux enhancement by using magnetic nanofluids and external magnetic fields

    International Nuclear Information System (INIS)

    By using the nanofluid as a working fluid, we can expect the enhancement in the flow boiling critical heat flux mainly due to the deposition of nanoparticles on the heat transfer surface. In this study, we suggest the magnetic nanofluid, or magnetite-water nanofluid, as a working fluid which is regarded as a controllable nanofluid, that is, nanoparticles or magnetite nanoparticles in a nanofluid can be controlled by an external magnetic field. Therefore, we can expect the advantages of magnetic nanofluid such as, i) control of nanofluid concentration to maintain nanoparticle suspension and to localize nanofluid concentration, and ii) removal of nanoparticle from nanofluid when we want. In this study, we focused on the investigation of flow boiling critical heat flux characteristics for the magnetic nanofluid. Series of experiments were performed under the low pressure and low flow conditions, and based on the experimental results; we can conclude that the use of magnetic nanofluid improves the flow boiling critical heat flux characteristics. This is mainly due to the deposition of magnetite nanoparticles on the heat transfer surface, which results in the improvement of wettability and re-wetting characteristics of heat transfer surface. Preliminary results of the magnetic field effects on the flow boiling critical heat flux would be presented also. (author)

  5. The Parable of the Boiled System Safety Professional: Drift to Failure

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    Jin, Miaomiao; Short, Michael

    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. 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.

  7. TRACG: Twenty years of collaboration between ENUSA and GE-HITACHI

    Energy Technology Data Exchange (ETDEWEB)

    Haces, J.; Trueba, M.; Garcia, J.; Barrera, J.

    2011-07-01

    TRACG is the GE Hitachi Nuclear Energy (GEH) proprietary version of the Transient Reactor Analysis Code. It is a best-estimate code for analysis of boiling eater reactors (BWR). Enusa has extensively contributed to the development of TRACG, applying this code to different scenarios and BWR plants: loss-of-coolant accident (LOCA), anticipated operational occurrences (AOO), instability events licensing of GNF fuel for Nordic plants, anticipated transients without scram (ATWS) reactivity insertion accidents (RIA), validation of the simulator for the Advanced BWR (ABWR) plant, the licensing of the TRACG based U. s. Nuclear Regulatory commission (NRC)-approved AOO and LOCA licensing methodologies, and in the licensing of the passively safe generation III+ Economic Simplified Boiling Water Reactor (ESBWR).

  8. Numerical study of subcooled boiling phenomena using a component analysis code, CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ba-Ro; Lee, Yeon-Gun [Jeju National University, Jeju (Korea, Republic of)

    2015-10-15

    In this study, a couple of subcooled boiling experiments at high- (> 10 bar) and low-pressure (near atmospheric pressure) conditions are analyzed using a three-dimensional thermal-hydraulic component code, CUPID. And then the analysis results compared with the results using MARS-KS code. Subcooled boiling experiments at high- and low pressure conditions are analyzed using a three dimensional thermal-hydraulic component code, CUPID. The predictions of the CUPID code shows good agreement with Christenses's data and Bartolomey's data obtained at high pressure conditions. Subcooled boiling is encountered in many industrial applications in the power and process industry. In nuclear reactors, under certain conditions, subcooled boiling may be encountered in the core. The movement of bubbles generated by subcooled boiling affect the heat transfer characteristics and the pressure drop of the system. Thus some experimental and analysis using safety codes works have been already performed by previous investigators. It has been reported that the existing safety analysis codes have some weaknesses in predicting subcooled boiling phenomena at low pressure conditions. Thus, it is required to improve the predictive capability of thermal-hydraulic analysis codes on subcooled boiling phenomenon at low-pressure conditions. At low pressure condition, the CUPID code generally is overestimated prediction of the void fraction. Thus, we did selected submodels in the heat partitioning model by sensitivity analysis. Selected submodels of M{sub c}ase 4 are Kocamustafaogullari and Ishii correlation model of active nucleate site density, N' and Fritz correlation model of bubble departure diameter, d{sub Bd} . And then, case 5 - 8 are reanalysis using submodels of M{sub c}ase 4. The calculated void fraction is compared the default CUPID code model to the modified CUPID code model. As a result, average void fraction error was reduced from 0.081 to 0.011 and 0.128 to 0.024, 0

  9. Co-boiling of NAPLs and water during thermal remediation: experimental and modeling study

    Science.gov (United States)

    Krol, M.; Zhao, C.; Mumford, K. G.; Sleep, B. E.; Kueper, B. H.

    2015-12-01

    The persistence of non-aqueous-phase liquids (NAPLs) in the subsurface has led to the development of several remediation technologies to address this environmental problem. One such group of technologies (in situ thermal treatment) uses heat to volatilize contaminants. Subsurface temperature measurements are often used to monitor progress and optimize contaminant removal. However, when NAPL and water are heated together, gas is created at a temperature lower than the boiling point of either liquid (co-boiling), which can affect temperature observations. To examine the effect of co-boiling on observed temperatures and NAPL mass removal, a series of heated laboratory experiments were performed using single and multi-component NAPLs. The experiments consisted of glass jars filled with a mixture of sand, water, and NAPL mixed to obtain an approximately uniform NAPL distribution within the jar. The experiments were heated from the outside and interior temperatures were measured using a thermocouple. The tests showed that local-scale temperature measurements are unreliable in indicating the end of co-boiling and may not indicate complete mass removal. This is because a well-defined co-boiling plateau does not exist when heating a multi-component NAPL and the temperature is dependent on the proximity of NAPL to the monitoring point. To further investigate temperature distributions and the potential to use gas production as a complementary indicator of NAPL removal, a 2D finite-difference mass transport model was used that incorporated heat transport, latent heat, phase change, and a multicomponent gas phase and used a macroscopic invasion percolation (MIP) model to simulate gas movement. Latent heat was calculated by multiplying specific latent heat, which is an intrinsic property of a substance, by the amount of liquid mass being vaporized and its incorporation into the model allowed for the simulation of co-boiling plateaus (during single component NAPL boiling). The

  10. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures Tsat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m2s and heat fluxes from 15 to 70 kW/m2, typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then utilised

  11. Expected reactivity effect of fuel channel coolant boiling in the Darlington NGS A reactor core

    International Nuclear Information System (INIS)

    We have developed a formalism for estimating the expected reactivity due to channel boiling in any reactor designed to have some quality in the channel. In applying this formalism to the Darlington NGS A equilibrium core, we calculate a value of 0.024 ± 0.003 mk at 100% power operation. In Darlington, the channel feeders are individually sized so that the coolant in each channel has some boiling on reaching the entrance to the reactor outlet header. (Hereafter called the 'ROH quality'). The design is such that when each channel is at its nominal time-averaged 100 percent power, the quality at the ROH should be just under 2%. The day-to-day variation of each channel's power around its time-averaged value (i.e., 'ripple') results in a corresponding variation in the quality and consequently in the reactivity due to boiling. Traditionally, fuel management codes such as SORO, FMDP, RFSP and OHRFSP use fuel properties generated by a lattice code such as POWDERPUFS or LATREP. These fuel properties are functions of fuel irradiation only, with all other core-varying input parameters to the lattice code held constant at core-averaged values. Recently, some work has gone into developing a Pt. Lepreau version of RFSP in which the fuel properties are functions of fuel temperature and coolant density as well as of fuel irradiation. This paper reports the results of a study which was undertaken to quantify the expected variation in core reactivity due to this day-to-day variation in channel power and channel boiling. It could then be determined whether the reactivity effect of this boiling is sufficient to justify the explicit representation of the fuel properties as a function of coolant density

  12. Numerical simulation of vapor bubble condensation in turbulent subcooled flow boiling

    International Nuclear Information System (INIS)

    Highlights: • The two-phase flow was modeled using VOF interface tracking method. • Bubble dynamics at high pressure, high velocity conditions was captured using LES. • Reverse bubble velocity trends occur for varying pressure with zero and nonzero bulk velocity. • Condensation rate increases with subcooling temperature and bulk velocity. • Bubble distortion increases with system pressure and bulk velocity. - Abstract: Subcooled flow boiling is a significant heat transfer regime in pressurized water reactors (PWRs) and also occurs in boiling water reactors (BWRs) prior to the onset of saturated nucleate boiling. In subcooled boiling, individual bubbles experience a life-cycle of nucleation from the wall, (possible) departure from the wall, and transport along the channel during condensation/collapse. The present paper focuses on the last stage of this life cycle. CFD analysis of a single bubble behavior in upward subcooled flow boiling was performed using volume-of-fluid (VOF) interface tracking method and large eddy simulation (LES) turbulence model. High pressure steam-water conditions at high flow velocity were simulated due to their relevance to the nuclear application. The numerical solutions were compared with terminal velocity correlations. Bubble behavior was investigated at different system pressure, subcooling temperature, bubble diameter, pipe diameter, and bulk fluid velocity, corresponding to bulk Reynolds number range of 0–410,000. The effects of these parametric variations on bubble rise velocity, deformation rate, trajectory, and distortion are presented. Results show that bubble rise velocity increases with bulk velocity, and bubble size, but decreases with increasing pipe diameter. A reverse bubble velocity trend with varying system pressure is observed with and without bulk velocity. Bubble condensation rate decreased with increasing system pressure, bubble size and pipe diameter, but increased with increasing subcooling temperature

  13. Numerical simulation of vapor bubble condensation in turbulent subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Owoeye, Eyitayo James, E-mail: msgenius10@ufl.edu; Schubring, DuWanye, E-mail: dlschubring@ufl.edu

    2015-08-15

    Highlights: • The two-phase flow was modeled using VOF interface tracking method. • Bubble dynamics at high pressure, high velocity conditions was captured using LES. • Reverse bubble velocity trends occur for varying pressure with zero and nonzero bulk velocity. • Condensation rate increases with subcooling temperature and bulk velocity. • Bubble distortion increases with system pressure and bulk velocity. - Abstract: Subcooled flow boiling is a significant heat transfer regime in pressurized water reactors (PWRs) and also occurs in boiling water reactors (BWRs) prior to the onset of saturated nucleate boiling. In subcooled boiling, individual bubbles experience a life-cycle of nucleation from the wall, (possible) departure from the wall, and transport along the channel during condensation/collapse. The present paper focuses on the last stage of this life cycle. CFD analysis of a single bubble behavior in upward subcooled flow boiling was performed using volume-of-fluid (VOF) interface tracking method and large eddy simulation (LES) turbulence model. High pressure steam-water conditions at high flow velocity were simulated due to their relevance to the nuclear application. The numerical solutions were compared with terminal velocity correlations. Bubble behavior was investigated at different system pressure, subcooling temperature, bubble diameter, pipe diameter, and bulk fluid velocity, corresponding to bulk Reynolds number range of 0–410,000. The effects of these parametric variations on bubble rise velocity, deformation rate, trajectory, and distortion are presented. Results show that bubble rise velocity increases with bulk velocity, and bubble size, but decreases with increasing pipe diameter. A reverse bubble velocity trend with varying system pressure is observed with and without bulk velocity. Bubble condensation rate decreased with increasing system pressure, bubble size and pipe diameter, but increased with increasing subcooling temperature

  14. Forced convective boiling heat transfer of water in vertical rectangular narrow channel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chong, E-mail: chenchong_2012@163.com; Gao, Pu-zhen, E-mail: gaopuzhen@hrbeu.edu.cn; Tan, Si-chao; Chen, Han-ying; Chen, Xian-bing

    2015-09-15

    Highlights: • Chen correlation cannot well predict the coefficient of rectangular channel. • Kim and Mudawar correlation is the best one among the Chen type correlations. • Lazarek and Black correlation predicted 7.0% of data within the ±30% error band. • The new correlation can well predict the coefficient with a small MAE of 14.4%. - Abstract: In order to research the characteristics of boiling flows in a vertical rectangular narrow channel, a series of convective boiling heat transfer experiments are performed. The test section is made of stainless steel with an inner diameter of 2 × 40 mm and heated length of 1100 mm. The 3194 experimental data points are obtained for a heat flux range of 10–700 kW/m{sup 2}, a mass flux range of 200–2400 kg/m{sup 2} s, a system pressure range of 0.1–2.5 MPa, and a quality range of 0–0.8. Eighteen prediction models are used to predict the flow boiling heat transfer coefficient of the rectangular narrow channel and the predicted value is compared against the database including 3194 data points, the results show that Chen type correlations and Lazarek and Black type correlations are not suitable for the rectangular channel very much. The Kim and Mudawar correlation is the best one among the 18 models. A new correlation is developed based on the superposition concept of nucleate boiling and convective boiling. the new correlation is shown to provide a good prediction against the database, evidenced by an overall MAE of 14.4%, with 95.2% and 98.6% of the data falling within ±30% and ±35% error bands, respectively.

  15. Visualization Study on High Heat Flux Boiling and Critical Heat Flux

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Satbyoul; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    In this study, an integrated visible and infrared-based experimental method is introduced to simultaneously measure the details of high-resolution liquid-vapor phase and heat transfer distributions on a heated wall. The dynamics and heat transfer at high heat flux boiling and critical heat flux were observed. The experiment was conducted in pool of saturated water under atmospheric pressure. There have been many studies to examine the physical mechanisms of nucleation boiling and critical heat flux over several decades. Several visible and infrared-based optical techniques for time-resolved high resolution measurements for liquid-vapor phase and heater surface temperature during boiling have been introduced to understand the characteristics and mechanisms of them. Liquid-vapor phase, temperature, and heat flux distributions on the heated surface were measured during pool boiling of water using the integrated total reflection and infrared thermometry technique. Qualitative examination of the data for high heat flux boiling and CHF was performed. The main contributions of this work are summarized below. The existence and behavior of dry patches lead the way toward CHF condition. Therefore, the mechanistic modeling of the CHF phenomenon necessarily needs to include the physical parameters related to dynamics of the large dry patch such as life time and size. In addition to the dynamic behavior of the dry patch, the thermal behavior of the hot patch is also important. Even though the dry area was rewetted, the stored thermal energy in the hot patch can be remained if the rewetting time is short and the subsequent dry patch is regenerated quickly.

  16. Heat Transfer of Single and Binary Systems inPool Boiling

    Directory of Open Access Journals (Sweden)

    Abbas J. Sultan

    2010-01-01

    Full Text Available The present research focuses on the study of the effect of mass transfer resistance on the rate of heat transfer in pool boiling. The nucleate pool boiling heat transfer coefficients for binary mixtures (ethanol-n-butanol, acetone-n-butanol, acetone-ethanol, hexane-benzene, hexane-heptane, and methanol-water were measured at different concentrations of the more volatile components. The systems chosen covered a wide range of mixture behaviors.The experimental set up for the present investigation includes electric heating element submerged in the test liquid mounted vertically. Thermocouple and a digital indictor measured the temperature of the heater surface. The actual heat transfer rate being obtained by multiplying the voltmeter and ammeter readings. A water cooled coil condenses the vapor produced by the heat input and the liquid formed returns to the cylinder for re-evaporation.The boiling results show that the nucleate pool boiling heat transfer coefficients of binary mixtures were always lower than the pure components nucleate pool boiling heat transfer coefficients. This confirmed that the mass transfer resistance to the movement of the more volatile component was responsible for decrease in heat transfer and that the maximum deterioration that was observed at a point was the absolute concentration differences between vapor and liquid phases at their maximum. All the data points were tested with the most widely known correlations namely those of Calus-Leonidopoulos, Fujita and Thome. It was found that Thome's correlation is the more representative form, for it gave the least mean and standard deviations

  17. Forced convective boiling heat transfer of water in vertical rectangular narrow channel

    International Nuclear Information System (INIS)

    Highlights: • Chen correlation cannot well predict the coefficient of rectangular channel. • Kim and Mudawar correlation is the best one among the Chen type correlations. • Lazarek and Black correlation predicted 7.0% of data within the ±30% error band. • The new correlation can well predict the coefficient with a small MAE of 14.4%. - Abstract: In order to research the characteristics of boiling flows in a vertical rectangular narrow channel, a series of convective boiling heat transfer experiments are performed. The test section is made of stainless steel with an inner diameter of 2 × 40 mm and heated length of 1100 mm. The 3194 experimental data points are obtained for a heat flux range of 10–700 kW/m2, a mass flux range of 200–2400 kg/m2 s, a system pressure range of 0.1–2.5 MPa, and a quality range of 0–0.8. Eighteen prediction models are used to predict the flow boiling heat transfer coefficient of the rectangular narrow channel and the predicted value is compared against the database including 3194 data points, the results show that Chen type correlations and Lazarek and Black type correlations are not suitable for the rectangular channel very much. The Kim and Mudawar correlation is the best one among the 18 models. A new correlation is developed based on the superposition concept of nucleate boiling and convective boiling. the new correlation is shown to provide a good prediction against the database, evidenced by an overall MAE of 14.4%, with 95.2% and 98.6% of the data falling within ±30% and ±35% error bands, respectively

  18. Film boiling heat transfer characteristics of sodium in droplet evaporation on heated tantalum

    International Nuclear Information System (INIS)

    For gaining background information on possible vapor explosion in a hypothetical core disruptive accident of liquid metal cooled fast breeder reactors, the experiment on the film boiling characteristics of sodium was conducted in association with Leidenfrost phenomenon. In a steel container filled with 1.0bar argon gas, sodium droplets were put on a heated disk and the behavior of droplets was observed through pyrex glass windows by a 35mm camera and a color videotape-recorder. A tantalum disk of 70mm dia. and 30mm height was induction-heated by an oil-cooled coil and a high frequency power supply of 20kHz and 30kW rating. The wall temperature of the disk was measured by a 1.6mm O.D. Ta-sheathed W-5%Re/W-26%Re thermocouple embedded beneath the disk. The experimental conditions were the initial droplet temperature and volume : 400-5000C and about 1.0cm3, the initial tantalum disk temperature : 1390-18900C. The heat flux was estimated from the volumetric reducing rate of droplet due to vaporization, based on photographic observation. The data plots of heat flux, though widely scattering, showed a decreasing trend with the wall superheat in the temperature range of 1390-16000C, while an increasing trend in the range of 1600-18900C. The former range suggests to correspond to the transition boiling region and the latter to the film boiling region. Thus, the minimum film boiling point was roughly estimated to be around 16000C and 45W/cm2. In the film boiling region the plots came slightly above the theoretical prediction. (author)

  19. Static flow instability in subcooled flow boiling in parallel channels

    Energy Technology Data Exchange (ETDEWEB)

    Siman-Tov, M.; Felde, D.K.; McDuffee, J.L.; Yoder, G.L. Jr.

    1995-04-01

    A series of tests for static flow instability or flow excursion (FE) at conditions applicable to the proposed Advanced Neutron Source reactor was completed in parallel rectangular channels configuration with light water flowing vertically upward at very high velocities. True critical heat flux experiments under similar conditions were also conducted. The FE data reported in this study considerably extend the velocity range of data presently available worldwide. Out of the three correlations compared, the Saha and Zuber correlation had the best fit with the data. However, a modification was necessary to take into account the demonstrated dependence of the Stanton (St) and Nusselt (Nu) numbers on subcooling levels, especially in the low subcooling regime.

  20. Overview on stability of natural-circulation-cooled boiling water reactors during start-up. An experimental and modeling analysis

    International Nuclear Information System (INIS)

    This paper provides an overview on numerical and experimental work focused on flashing-induced instabilities. These instabilities may occur in natural circulation two-phase systems when operated at low pressure and low power. Therefore they are of special interest for the start-up phase of natural circulation Boiling Water Reactors. The work presented in this paper has been performed within the framework of the NACUSP project (European-Union Fifth Framework Program). Experiments were carried out on a steam/water natural circulation loop (CIRCUS), built at the Delft University of Technology. Information was gained on the characteristics of the flow oscillations and on the void fraction production during flashing in stationary and transient conditions. A 3-D flow-pattern visualization was achieved by means of advanced instrumentation, namely wire-mesh sensors. On the basis of the experimental results, an assessment of existing drift-flux models was performed for flashing flow. The most suitable drift-flux model was implemented in the 4-equations two-phase model FLOCAL, developed at the Forschungszentrum Rossendorf (FZR, Germany). The model allows for the liquid and steam to be in thermal non-equilibrium and, via drift-flux models, to have different velocities. A detail comparison between simulations and experiments is reported. (author)