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Sample records for abwr advanced boiling

  1. Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

    European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

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

    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.

  3. Main control panel design and simulator in advanced boiling water reactor [ABWR

    The ABWR type main control panel has been developed to enhance the reliability of plant monitoring and operation. This panel consists of large display panels and a compact main console, which enables operators to work from their seated position. Primary plant and system status information is presented on large display panels so that the information can be shared by the entire operating crew. This panel is applied to Kashiwazaki-Kariwa Nuclear Power Station Unit 6 which is scheduled to begin commercial operation in 1996. The ABWR type main control panel is quite different from the conventional one. Consequently the full-scope ABWR simulator is under construction, which will be completed in 1994, so that the operators will be trained sufficiently. (author)

  4. Outline of advanced boiling water reactor

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

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

    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)

  6. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    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...

  7. Advanced boiling water reactor

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  8. Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

    This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

  9. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    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...

  10. The development of ABWR

    The first Advanced Boiling Water Reactor (ABWR) started commercial operation as Tokyo Electric Power Company's (TEPCO) Kashiwazaki-Kariwa Nuclear Power Station Unit No.6 (K-6) in November 1996 and its sister Unit No.7 (K-7) in July 1997. The ABWR was developed to achieve higher reliability and safety margin while improving overall operability and economics. To achieve these goals, the optimal Boiling Water Reactor (BWR) technologies had been studied, tested and were finally adopted into the ABWR design. These technologies were called 'First of a Kind' and include the Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), and integrated digital Instrumentation and Control System (I and C). Intensive development study, confirmation tests and verification tests were conducted by the plant equipment suppliers, electric utilities, and government agencies. During plant construction, the system and equipment functions and operational characteristics were confirmed through various tests. Before the start of the commercial operation, it was confirmed that these first ABWR units met all the design goals that had been established for the next generation of nuclear power plants. Both units have now completed the first fuel cycle of operation without any unplanned plant shut down and have achieved very high availability factors. This paper describes the development and construction of these first ABWR units in the world. (author)

  11. Development of Innovative Construction Technologies for ABWRs

    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

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

    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.

  13. U.S. licensing process and ABWR certification

    As part of the US ALWR Program to revitalize the nuclear option through the integration of government/utility/industry efforts, GE undertook the role of applying for certification for its latest product line, the advanced boiling water reactor (ABWR), under the US ABWR certification program. The ABWR utilizes proven BWR technology. Two units are under construction in Japan. The ABWR design is an essentially complete plant. Initial application for certification was in 1987 under Part 50. GE reapplied in late 1991 under the newly promulgated Part 52. Following 7 years of intensive interactions with the NRC and ACRS, GE was awarded the first FDA under Part 52. The commission initiated rulemaking by publishing the proposed ABWR Certification Rule in the Federal Register in early 1995. Certification is anticipated mid-1996. In addition to applying for ABWR certification, GE undertook the ABWR First-of-a-Kind (FOAKE) contract in June 1993 to perform a detailed design of the ABWR for application in the US. The purpose of this program is to develop the ABWR design to the point where a reliable construction cost estimate can be made

  14. Advanced plant engineering and construction of Japanese ABWRs

    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)

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

    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)

  16. Steam turbine and generator system for ABWR

    The advanced boiling water reactor (ABWR), with a number of superior characteristics including high reliability and large capacity, has been developed. The first ABWR units have been realized as Units No. 6 and 7 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station. Based on its considerable experience in the construction and maintenance of nuclear steam turbine and generator systems, Toshiba has developed the world's largest class steam turbine as well as generator systems with high efficiency, high reliability, small turbine buildings and a short construction period, and has been constructing Kashiwazaki-Kariwa Unit No. 7 jointly with General Electric Co. of the United States. We have been developing additional techniques to further improve the efficiency and maintainability of the steam turbine and generator system of the next ABWR plant, based on techniques that have been verified with fossil-fuel power plants. (author)

  17. Digital control application for the advanced boiling water reactor

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

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

    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

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

    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

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

    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.

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

    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

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

    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.

  3. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    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...

  4. ABWR certification work brings US licensing stability nearer

    The Advanced Boiling Water Reactor (ABWR) is now approaching Final Design Approval by the US Nuclear Regulatory Commission (NRC) and will then proceed on to the certification phase of the NRC's new standard plant licensing process. Successful completion of this will usher in a new era of standardization and reactor licensing stability in the US. (author)

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

    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)

  6. Human factors evaluation of the new control panels for ABWR

    The new control panels for the Advanced Boiling Water Reactor (ABWR) are characterized by the use of a large display panel, hierarchical configuration of alarm functions, the enlarged scope of automation function, a compact main console and operation by CRTs and FDs (flat displays) with touch sensitive screens. For evaluation of the technologies newly introduced into the ABWR control room, data from the ABWR simulator were compared with the data from the existing type simulator. Usability evaluation was performed by analyzing questionnaire data, interview results, simulator log, audio visual data and mental workload data. The following results were obtained for the new human-machine interface: a) The large display panels significantly reduced operators workload and improved crew communication. b) Under the condition of multiple failures with all CRTs unavailable, it was possible to shut down the plant safety by using flat displays and large display panels

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

    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

  8. Cost reduction and safety design features of ABWR-II. Annex 5

    The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

  9. Development of medium sized ABWRs (ABWR-600, ABWR-900)

    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 600 MWe class reactor (ABWR-600) which is appropriate for a distributed power supply. The next step was to develop an uprating concept to extend this output scale to the 900 MWe class reactor (ABWR-900) based on the rationalized technology of the ABWR-600 for further cost savings. (author)

  10. Development of medium sized ABWR (ABWR-600, ABWR-900)

    Hitachi has been developing the Medium sized ABWR as a power source that features flexibility to various market needs, such as minimized capital risks, timely return on a capital investment, etc. Basic design policy of the Medium sized ABWR is 1) to be based on ABWR design with superior experiences in construction and operation, 2) to utilize standard BWR fuel adopting proven technology, 3) to achieve rationalized design by suitably utilizing the key components developed for large sized reactor. Development of the Medium sized ABWR has been working to proceed step by step. The first step was to design an output scale of the 600 MWe class (ABWR-600), which is appropriate for a distributed power-supply. The next step we start developing of the up-rated to an output scale of the 900 MWe class (ABWR-900) based on ABWR-600 rationalized technology for further cost saving. (author)

  11. Fuzzy logic control of water level in advanced boiling water reactor

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

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

    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)

  13. Modeling pressure suppression pool hydrodynamics in the ABWR containment

    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.

  14. Innovative techniques applied to ABWR project engineering

    General Electric's (GE) Advanced Boiling Water Reactor (ABWR) project is characterised by the use of new production methods and tools, a document configuration system that was defined from the outset and wide-ranging, smooth communications. The project also had a large number of participating companies from different cities in the US (San Jose, San Francisco, Kansas City, Washington), Mexico (Veracruz) and Spain (Madrid). One of the basic requirements applicable to advanced nuclear power plant projects is the need for an Information Management System (IMS) which shall be valid for the entire life of the plant, which means that all the documentation must be available in electronic format. The basic engineering tool for the ABWR project is POWRTRAK, a computer application developed by Black and Veatch (B and V). POWRTRAK comprise a single database, in which each datum is stored in only one place and used in real time. It consists of various modules, some of which are associated with technical data and the generation of diagrams (CASES, application used to generate piping and instrumentation, logic and electric wiring diagrams), three-dimensional electronic mock-up, planning, purchasing management, etc. GE adapted the Odesta Document Management System (ODMS) commercial application to its documentation file/control needs. In this system all the documentation produced in the project is filed in both native and universal formats (PDF). (Author)

  15. 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

    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.

  16. ABWR construction experience in Japan

    The construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and 7 (K6 and 7) owned by Tokyo Electric Power Company (TEPCO), have been in service since November 1996 and September 1997 respectively, and are the first ABWRs in the world. Hitachi, Ltd. and Toshiba Corporation took part in the projects and achieved safety construction in 52- month schedule from rock inspection to turnover in cooperation with TEPCO and other contractors with broad experience in BWRs. Taking advantage of experience of the Kashiwazaki construction, the ABWR construction methods realize to satisfy with requests from electric power companies and to fit site conditions. The Chubu Electric Power Company's Hamaoka Nuclear Power Station Unit No. 5 (Hamaoka 5) and Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2), which are ABWRs, completed rock inspection in May 2000 and June 2001 respectively, and are under construction. The ABWR construction methods are proved their validity through four actual construction projects. Safe constructions with the highest quality have been advanced thus further aiming at the introduction of the latest technologies and contributing to reduce plant construction costs and also to shorten schedule. (author)

  17. Design certification program of the simplified boiling water reactor

    General Electric (GE), the US Department of Energy, the Electric Power Research Institute (EPRI), and utilities are undertaking a cooperative program to enable advanced light water reactor (ALWR) designs to be certified by the US Nuclear Regulatory Commission (NRC). GE is seeking to certify two advanced plants; the Advanced Boiling Water Reactor (ABWR) and the Simplified Boiling Water Reactor (SBWR). Both plants use advanced features that build on proven BWR technology

  18. Participation in the ABWR Man-Machine interface design. Applicability to the Spanish Electrical Sector

    Project coordinated by DTN within the advanced reactor programme. Participation in the design activities for the Advanced Boiling Water Reactor (ABWR) man-machine interface was divided into two phases: Phase I: Preparation of drawings for designing, developing and assessing the advanced control room Phase II: Application of these drawings in design activities Participation in this programme has led to the following possible future applications to the electrical sector: 1. Design and implementation of man-machine interfaces 2. Human factor criteria 3. Assessment of man-machine interfaces 4. Functional specification, computerised operating procedures 5. Computerised alarm prototypes. (Author)

  19. Effects of lower plenum flow structure on core inlet flow of ABWR

    The evaluation of coolant flow structure at a lower plenum of an advanced boiling water reactor (ABWR) in which there are many structures is very important in order to improve generating power. Although the simulation results by CFD (Computational Fluid Dynamics) codes can predict such complicated flow in the lower plenum, it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of the CFD codes. In the model of the lower plenum, we measured velocity profiles with LDV and PIV. And differential pressure of constructed model is measured with differential pressure instrument. It was identified that the velocity and differential pressure profiles also showed the tendency to be flat in the core inlet. Moreover, vortexes were observed around side entry orifice by PIV measurement. (author)

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

    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.

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

    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.)

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

    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)

  3. Alarm system for ABWR main control panels

    TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ''Alarm System for ABWR Main Control Panels''. (author). 5 refs, 8 figs, 1 tab

  4. Improvements in boiling water reactor designs and safety

    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

  5. An ABWR water chemistry control design concept for low radiation exposure and the operating experience at the first ABWR

    No. 6 and 7 units of the Kashiwazaki-Kariwa Nuclear Power Station are the first advanced boiling water reactors (ABWR). Unit No. 6 (K6) started commercial operation in November 1996 and unit No. 7 (K7), which is identical to unit 6, began operation in July 1997. Both of them have now finished three cycles of operation. In the design stage of the ABWR, specifications for material selection were based on a detailed review of corrosion product mass balance from the standpoint of radiation source reduction. Improved cobalt specifications were applied to the stainless steel of the feed water heater tube and corrosion resistant steel was used in the turbine system. K6 and K7 adopted different water chemistry controls to reduce the dose rate during plant operation. K6 used extremely low iron crud control in the feed water and K7 used improved Ni/Fe control. As the result of these challenges, both of them attained the design target of less than 0.36 mSv/year radiation exposure. (authors)

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

    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

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

    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

    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. Negative pressure difference evaluation of Lungmen ABWR containment by using GOTHIC

    Highlights: • The Wetwell-to-Drywell Vacuum Breakers (WDVBs) can mitigate the negative pressure difference. • The performance of the WDVBs of the Lungmen ABWR containment is analyzed. • There is sufficient margin unless only one of eight WDVBs is operable. • The WDVBs shall be closed to avoid containment pressurization during accident. • The current requirement of the Lungmen plant ensures the containment safety. - Abstract: Negative pressure difference of the pressure suppression containments means that the wetwell pressure is greater than the drywell pressure and is an undesired effect to the containment safety. The Wetwell-to-Drywell Vacuum Breakers (WDVBs) are check valves used to mitigate this adverse effect. The Lungmen Nuclear Power Plant in Taiwan is a twin-unit Advanced Boiling Water Reactor (ABWR) plant. There are totally 8 WDVBs in the ABWR containment. In this study, a GOTHIC model is developed to evaluate the negative pressure difference of the Lungmen plant. Sensitivity study on the operable WDVB number is also performed. The results show that there is sufficient margin against the negative pressure difference for the ABWR containment. The design value will not be exceeded unless only one WDVB is operable. Furthermore, the effect of the WDVB leakage on the drywell pressure is evaluated by performing the short-term containment analyses. The peak drywell pressure will be greater than the design value with leakage area greater than 50% of one WDVB area. Based on the results of this study, the current requirement that all WDVBs shall be closed and operable during normal operation ensures the Lungmen containment safety

  10. Boiling heat transfer modern developments and advances

    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

  11. Boils

    ... 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. Study of Pu consumption in Advanced Light Water Reactors

    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

  13. Boils

    ... or recurrent boils, which are usually due to Staph infections. The bacteria are picked up somewhere and then ... version of boils is folliculitis . This is an infection of hair follicles, usually with Staph bacteria. These often itch more than hurt. The ...

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

    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

  15. Neutronic challenges of advanced boiling water reactor designs

    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)

  16. Features of ABWR operator training with a full-scope simulator

    Many innovations have been incorporated into the Advanced BWR (ABWR) type control panels. In the BWR Operator Training Center (BTC), we started ABWR operator training using an ABWR full-scope simulator prior to the first ABWR plant's commercial operation. In consideration of the features of the ABWR type control panels, BTC has been conducting ABWR operator training focusing on the following 2 points; (1) Operator training reflecting the differences in the Human-Machine Interface (HMI). The new HMI devices which have the touch-operation function were introduced. These devices have higher operability, however, they require new operational skills. We planned the training program so that operators can fully acquire these skills. Also the compact main console and the new HMI devices made it relatively difficult for the operator crews to grasp visually what an operator was doing. We provide the training to have proper communication skills, and check trainees' operation using monitoring systems for simulator training. (2) Operator training responding to the expanded operation automation system. The scope of the automation system was expanded to reduce the operators' burden. We provide the training to improve the trainees' competence for 'operation and monitoring' suitable to both manual and automatic operational modes. (author)

  17. Further improvement of human-machine interface for ABWR main control room

    Tokyo Electric Power Company (TEPCO) has developed main control room panels based on progress in C and I technology. ABWR type main control room panels (ABWR MCR PNLs) are categorized as third generation type domestic BWR MCR, that is, they are were developed step by step based on operating experience with the first and the second generation BWR. ABWR type main control room panels were applied to Kashiwazaki-Kariwa Nuclear Power Station Units Number 6 and 7 (K-6/7) for the first time. K-6/7 are the first advanced BWR (ABWR), which started commercial operation in November 1996 and July 1997, respectively. The concept of ABWR MCR design was verified through wooden mock-up panels, start-up tests and commercial operation. Though the K-6/7 design has borne fruit, we are planning to refine and standardize the design based on the following concepts: to maintain the plant operation and monitoring style of ABWR MCR PNLs; to introduce brand-new HMI technology and devices; to incorporate operators' advice in the design. This paper outlines the features and improvements of the K6/7 MCR PNLs design. (author)

  18. Prediction of a subcooled boiling flow with advanced two-phase flow models

    Highlights: ► In this study, advanced two-phase flow models were examined to enhance the prediction capability of subcooled boiling flows for the CFD code. ► They consist of Sγ bubbles size, new wall boiling and two-phase logarithmic wall function models. ► The benchmark calculation confirms that advanced two-phase flow models show good prediction results. - Abstract: Prediction of bubble size which governs interfacial transfer terms between the two phases is of importance for an accurate prediction of the subcooled boiling flow. In the present work, a mechanistic bubbles size model, Sγ was examined to enhance the prediction capability of subcooled boiling flows for the CFD (computational fluid dynamics) code. In addition to this, advanced subcooled boiling models such as new wall boiling and two-phase logarithmic wall function models were also applied for an improvement of energy partitioning and two-phase turbulence models, respectively. The benchmark calculation against the DEBORA subcooled boiling data confirms that the Sγ bubble size model with the two advanced subcooled boiling models shows good prediction results and is applicable to the wide range of flow conditions that are expected in the nominal and postulated accidental conditions of a nuclear power plant.

  19. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

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

    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.

  1. Commercial operation and outage experience of ABWR at Kashiwazaki-Kariwa units Nos. 6 and 7

    Kashiwazaki-Kariwa Nuclear Power Station Units Nos. 6 and 7, the world's first ABWRs (Advanced Boiling Water Reactor), started commercial operation on November 7, 1996 and July 2, 1997, respectively, and continued their commercial operation with a high capacity factor, low occupational radiation exposure and radioactive waste. Units 6 and 7 were in their 3rd cycle operation until 25th April 1999 and 1st November 1999, respectively. Thermal efficiency was 35.4-35.8% (design thermal efficiency: 34.5%) during these period, demonstrating better performance than that of BWR-5 (design thermal efficiency: 33.4%). Nos. 6 and 7 have experienced 2 annual outages. The first outage of unit No. 6 started on November 20, 1997 and was completed within 61 days (including 6 New Year holidays), and the second outage started on March 13, 1999 and was completed within 44 days. The first annual outage of unit No. 7 started on May 27, 1998, earlier than it would normally have been, to avoid an annual outage during the summer, and was completed within 55 days, and the second outage started on September 18th, 1999 and was completed within 45 days, All annual outages were carried out within a very short time period without any severe malfunctions, including newly designed ABWR systems and equipment. As the first outage in Japan, 55 days is a very short period, despite the fact that the Nos. 6 and 7 are the first ABWRs in the world and the largest capacity units in Japan. The total occupational radiation exposure of No. 6 was 300 man-mSv (1st outage) and 331 man-mSv (2nd outage). That of Unit 7 was 153 man-mSv (1st outage) Those of unit No. 6 were at the same level as those of unit No. 3, which is the latest design 1100MW(e) BWR-5. That of unit No. 7 was the lowest ever at Kashiwazaki-Kariwa nuclear power station. The drums of radioactive waste discharged during the annual outage numbered 54 (1st outage) for No. 6 and 62 (1st outage) for No. 7, which was less than the design target of 100

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

    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

  3. Hitachi's experience and achievements in ABWR construction

    Hitachi, Ltd. (Hitachi) is taking charge of the turbine island of Chubu Electric Power Company's Hamaoka nuclear power plant Unit No. 5 (Hamaoka 5), and all facilities of Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2). Hitachi has achieved shorter schedules and lower costs in the two projects with the following construction methods. First, Hitachi applied modularization and the area-by-area method. These methods enable us to undertake building and installation work in parallel. Second, the floor packaging construction method enables us to finish installation in every area earlier than conventional methods. Moreover, manpower can be distributed evenly. This paper reports the application of these advanced construction methods to the Hamaoka 5 and Shika 2 projects. Installations such as RCCV modules and condenser blocks are presented in this paper. In addition, Hitachi is using 3D-CAD engineering and our own construction management system to optimize the planning of construction procedures and the installation of equipment in these projects. Thus Hitachi is always striving to improve the reliability, safety and economy of ABWR construction technologies. (author)

  4. Study on advanced reinforced concrete containment for the next generation boiling water reactors

    The paper presents the study results about advanced reinforced concrete containment (ARCCV) for the next generation BWRs. The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The ARCCV is based on the current reinforced concrete containment vessel (RCCV) which is used for the current ABWR. There are three major improvements from the current RCCV. These improvements enables the reduction of the R/B volume and improvement of safety simultaneously. The first improvement of the ARCCV is the separation between the upper DW and the lower DW. This contributes to the reduction of the nitrogen and hydrogen transferred to the WW air space at an accident. The pressure increase by the compression in the WW air space can be minimized and hence the reduction of the peak pressure of the containment dramatically regardless of the power increase to) 1,700 MWe. The second improvement of the ARCCV is the adoption of the advanced passive cooling system (APCS). It can minimize the necessity of the containment venting at severe accidents. The third improvement of the ARCCV is the raised S/P. Most of the ECCS pumps can be installed under the raised S/P. This can reduce the area of the lowest floor of the R/B and hence the reduction of the R/B volume. The above three improvements enables the cost saving of the R/B and safety improvements for both DBA LOCA and severe accidents dramatically. (author)

  5. Post-test analysis of a 1:10-scale top slab model of ABWR/RCCV subjected to internal pressure

    Construction of the first Advanced Boiling Water Reactor (ABWR) in Japan employing a reinforced concrete containment vessel (RCCV) was started in 1991. As RCCV itself is the first structure of its kind in Japan, thorough verification tests have been performed. This paper presents the results of simulation analysis of the Top Slab partial model of the RCCV subjected to internal pressure beyond design load. The Top Slab portion is complicated, being composed of a flat Top Slab, cylindrical wall and fuel pool girders, that its simulation analysis requires the evaluation of nonlinear structural behavior of reinforced concrete members due to membrane, bending and shear forces. This paper reports that Finite Element analysis with 3-D solid elements has given a good quantitative agreement between experimental and analysis results with respect to deformation, failure load and each nonlinear behavior. (orig.)

  6. Post-test analysis of a 1:10-scale top slab model of ABWR/RCCV subjected to internal pressure

    The first Advanced Boiling Water Reactor (ABWR) employing a reinforced concrete containment vessel (RCCV) has started its construction in 1991 in Japan. As RCCV itself is the first structure of its kind in Japan, thorough verification tests have been performed. This paper presents the results of simulation analysis of the top slab partial model of the RCCV subjected to internal pressure beyond design load. Top slab portion is so complicated, composed of flat top slab, cylindrical wall and fuel pool girders, that its simulation analysis requires the evaluation of nonlinear structural behavior of reinforced concrete members due to membrane, bending and shear forces. This paper reports that finite element analysis with 3-dimensional solid approach has given a good agreement quantitatively between experimental and analysis results with respect to deformation, failure load and each nonlinear behaviors

  7. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    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...

  8. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    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...

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

    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

  10. Advanced analytical techniques for boiling water reactor chemistry control

    The analytical techniques applied can be divided into 5 classes: OFF-LINE (discontinuous, central lab), AT-LINE (discontinuous, analysis near loop), ON-LINE (continuous, analysis in bypass). In all cases pressure and temperature of the water sample are reduced. In a strict sense only IN-LINE (continuous, flow disturbance) and NON-INVASIVE (continuous, no flow disturbance) techniques are suitable for direct process control; - the ultimate goal. An overview of the analytical techniques tested in the pilot loop is given. Apart from process and overall water quality control, standard for BWR operation, the main emphasis is on water impurity characterization (crud particles, hot filtration, organic carbon); on stress corrosion crackling control for materials (corrosion potential, oxygen concentration) and on the characterization of the oxide layer on austenites (impedance spectroscopy, IR-reflection). The above mentioned examples of advanced analytical techniques have the potential of in-line or non-invasive application. They are different stages of development and are described in more detail. 28 refs, 1 fig., 5 tabs

  11. Power distribution control within the scope of the advanced nuclear predictor for boiling water reactors

    In boiling water reactors the Advanced Nuclear Predictor (FNR) has proved to be a valuable tool in improving plant operating efficiency. The system is described in its main features and capabilities. As a logical extension, a power distribution control system has been developed, based on a reduced but accurate core model, which in itself can be used for fast prediction of core states. The system provides prediction of optimal operating strategies as well as on-line control, observing all constraints imposed on the permissible operating region. (orig.)

  12. Novelties in design and construction of the advanced reactors

    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

  13. Self-Sustaining Thorium Boiling Water Reactors

    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...

  14. Verification of advanced methods in TARMS boiling water reactor core management system

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site Boiling Water Reactor (BWR) core operation management system. It covers almost all the functional requirements to the current process computer to increase on-site core management capability, capacity factors, thermal margins, fuel reliability, and so on, by supporting application functions for monitoring the present core power distribution, and for aiding site engineers in making the core operation plans, by predicting future core performance. It is based on a three dimensional, 1.5 energy group, coarse mesh nodal diffusion theory code ''LOGOS02'', and includes advanced methods to increase the accuracy of core power distribution calculations as well as a local peaking factor calculation method by which the effect of neighboring nodes on intra-nodal power distribution can be considered. TARMS has been installed in eight BWR plants and was verified to be an effective BWR core operation management tool. This paper describes its advanced methods and the results of verifications with actual plant data. (author). 3 refs, 6 figs

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

    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

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

    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.

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

    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

  18. A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

    In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and

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

    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.

  20. The status of development activities of ABWR-2

    This paper reports the current status of development activities for ABWR-II which is the next generation design based on ABWR. Six Japanese BWR utilities and three BWR plant venders, General Electric Company, Hitachi Ltd. and Toshiba Corporation, have jointly been promoting the program. The program has been progressing for a decade. In Phase I (1991-92), future technologies were discussed and several plant concepts were studied. In Phase II (1993-95), in order to establish a reference reactor concept, key design features were selected. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements have been done in order to satisfy the design requirements well. The start of commercial operation of ABWR-II is estimated to be in the late 2010 when replacements of the first generation of operating nuclear power plants are expected to start. Larger power output and some passive safety features are adopted in order to make ABWR-II competitive in power generation cost and attain public confidence on safety. The advantages of a large fuel bundle design are also discussed here. (author)

  1. Development, operating experience and future plan of ABWR in Japan

    From 1974 to 1978 nuclear power plants' capacity factor decreased to a low level at the Tokyo Electric Power Company (TEPCO) for stress corrosion cracking of stainless steal pipe, thermal fatigue crack of feed water spurger and vibration trouble of LPRM, etc. So, for improving that situation, we gathered the most useful BWR technology in the world at that time and we started the ABWR project to develop an ideal BWR. The goals of this development were: (1) Enhanced safety and reliability; (2) Reduced occupational radiation exposure and radioactive waste; (3) Enhanced operability and maneuverability; and (4) Improved economy. Significant improvements were induced by the adoption of the reactor internal pump, fine motion control rod drive mechanism, integral type reinforced concrete containment vessel and digital control system. After about 10 years R and D, we succeeded in the development of the compact and economic plant, ABWR (1356MW(e)) whose building volume had been reduced to 76%, compared with the previous BWR (1100MW(e)). The construction of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6, which is the first ABWR plant in the world, was started September 1991, and that of unit No. 7 was started in February 1992. We used effective construction methods, such as large block construction method and all weather construction method, to shorten the construction duration from first concrete to commercial operation start to 51.5 months. The commercial operation start of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6 was November 1996 and that of unit No. 7 was July 1997. After that, unit No. 6 had three times the annual outage and unit No. 7 had two times the annual outage. Both plants have good operating experiences. From commercial operation start to end of August 2000, the capacity factor of unit No. 6 is 86.4% and that of unit No. 7 is 86.9%. Now, in Taiwan, ABWR plants are under construction at Lungmen unit No. 1 and 2. In Japan, ABWR plants are under

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

    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 ...

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

    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)

  4. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

  5. Boils (Furunculosis)

    ... 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 ...

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

    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

  7. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    organizations from 7 countries, SMART, integrated reactor, developed by Korea Atomic Energy Research Institute, Republic of Korea; CAREM, Argentina integrated reactor; MRX, integrated reactor, developed by Japan Atomic Energy Research Institute; UNITERM, NPP with integrated reactor, development by Research and development institute of power engineering (NIKIET), Russian Federation; AHEC-80, Russian NPP with integrated reactor, developed by OKB Mechanical Engineering (OKBM), Nizhny Novgorod, Russia. Moreover, following Boiling Water Reactor (BWR) projects have been subjected to the system comparative analysis. 1) Large Sized Reactors: ABWR, developed by Hitachi, Ltd, Japan, Toshiba Corporation, Japan ? G.E. USA; BWR-90, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; BWR-90+, developed by Nuclear Systems Division, ABB Atom AB, Vasteras, Sweden; SWR 1000, developed by Siemens Corporation, Germany; ESBWR, developed by General Electric Company, USA. 2) Medium Sized Reactors: SBWR, developed by General Electric Company, USA; HSBWR, developed by Hitachi Company, Ltd. 3) Small Sized Reactors: SSBWR, developed by Hitachi, Ltd, Japan; VK-300, BWR reactor, developed by Research and Development Institute of Power Engineering (NIKIET), Russia. Some data on the analysis of the condition and prospects of energy production and energy consumption, stations and networks in Kazakhstan are given. According to this analysis of nuclear power plants of average and low power are considered to be the most appropriate to construction in Kazakhstan. Recommendations on a choice of the most safe, reliable and economically competitive reactors have been made among the above-mentioned ones PWR, WWER and BWR for construction in Kazakhstan

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

    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

  9. Neutronic evaluation of a non-fertile fuel for the disposition of weapons-grade plutonium in a boiling water reactor

    A new non-fertile, weapons-grade plutonium oxide fuel concept is developed and evaluated for deep burn applications in a boiling water reactor environment using the General Electric 8x8 Advanced Boiling Water Reactor (ABWR) fuel assembly dimensions and pitch. Detailed infinite lattice fuel burnup results and neutronic performance characteristics are given and although preliminary in nature, clearly demonstrate the fuel's potential as an effective means to expedite the disposition of plutonium in existing light water reactors. The new non-fertile fuel concept is an all oxide composition containing plutonia, zirconia, calcia, and erbia having the following design weight percentages: 8.3; 80.4; 9.7; and 1.6. This fuel composition in an infinite fuel lattice operating at linear heat generation rates of 6.0 or 12.0 kW/ft per rod can remain critical for up to 1,200 and 600 Effective Full Power Days (EFPD), respectively, and achieve a burnup of 7.45 x 1020 f/cc. These burnups correspond to a 71--73% total plutonium isotope destruction and a 91--94% destruction of the 239Pu isotope for the 0--40% moderator steam void condition. Total plutonium destruction greater than 73% is possible with a fuel management scheme that allows subcritical fuel assemblies to be driven by adjacent high reactivity assemblies. The fuel exhibits very favorable neutron characteristics from beginning-of-life (BOL) to end-of-life (EOL). Prompt fuel Doppler coefficient of reactivity are negative, with values ranging between -0.4 to -2.0 pcm/K over the temperature range of 900 to 2,200 K. The ABWR fuel lattice remains in an undermoderated condition for both hot operational and cold startup conditions over the entire fuel burnup lifetime

  10. Start of ABWR to move. Trial operation of Kashiwazaki-Kariwa Nuclear Power Station, No. 6 Reactor (from equipments installation to start of commercial operation)

    Development of ABWR have been reflected many lessons based on advanced technology and passed experiences on operation of BWRs in the world and promoted under aims of 1) improvement of safety and reliability, 2) reduction of irradiation received by workers, 3) reduction of amounts of radioactive wastes, 4) improvement of operability, and 5) improvement of economics. As technical characteristics of the ABWR, it can be shown that recirculation pump of coolant is installed within the reactor as an internal pump, that channels in control rod driver is improved and that storing container made of steel reinforced concrete is adopted. For these new design technology and new design equipments, given performance, function and design validity were confirmed by using for preparing and constructing actually Kashiwazaki-Kariwa nuclear power station, Nos. 6 and 7 reactors, the first ABWR plants in the world and by executing various inspections and tests. In this paper, trial operation results containing the other starting operational test, discontent state during trial operation and so forth were introduced. (G.K.)

  11. GE's advanced nuclear reactor designs

    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

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

    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

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

    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

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

    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

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

    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)

  16. Self-Sustaining Thorium Boiling Water Reactors

    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.

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

    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

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

    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)

  19. Oscillate Boiling

    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...

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

    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)

  1. Human factors verification and validation of the advanced nuclear plant control room design

    The GE Advanced Boiling Water Reactor (ABWR) design has implemented the applicable human factors engineering (HFE) principles in the design of human-system interfaces (HSI). The ABWR uses unique features such as large mimic and touch-screen technology to present plant overviews and system operating details to the control room operating staff. The HSI designs, both in the console panels and the software generated graphical user interfaces, have been developed and evaluated using HFE guidelines. In addition to HFE guidelines reviews performed during design and implementation, broader reviews have been performed under the HFE Verification and Validation Implementation Plan (HFE V and VIP). Based upon the NUREG-0711, Nuclear Regulatory Commission (NRC) HFE Program Review Model (HFE PRM) (Reference 1), the HFE V and VIP, hereafter also referred to as V and V, has provided feedback during the various phases of design, implementation, and integration of the HSI. As one of the ten elements of the HFE PRM, the V and V activities reaffirm that the design of the HSI conforms to the HFE design principles and that the plant operating staff in the control room can perform their assigned tasks. This rigorous HFE V and V process is now being applied in the implementation of the ABWR design for Taiwan Power Company's Lungmen Power Station. Two 1350 MWe ABWR units are currently under construction at Lungmen. The HFE V and V ensures that the process for the design is compliant with the HFE principles. 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. These HFE V and V activities, performed in three phases, ensures that the necessary displays, control, and alarms are provided to support the identified personnel tasks. The HFE V and V also checks to determine that the design of each identified component is compliant with the HFE principles. The V and V ensures

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

    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

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

    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.

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

    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

    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

  6. The steam-line break inside containment LOCA transient analysis of Lungmen ABWR with TRACE

    The US NRC is developing TRACE (TRAC/RELAP Advanced Computational Engine), a new thermal-hydraulic code for safety analysis of nuclear power plants. Under the terms of the CAMP (Code Applications and Maintenance Program) contract, the US authority joined most of countries who own nuclear power plants, like Taiwan, together to participate in the research and application of TRACE code. TRACE is a modernized code with the capability to simulate the reactor system and model the thermal-hydraulic phenomena in three-dimensional space. Instead of those out-of-date codes like TRAC and RELAP, TRACE has become the NRC's flagship thermal-hydraulic analysis tool. One of the features of TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation capability than the other old codes, especially for events like LOCA (Loss of Coolant Accident). Following in Japan's footstep, Taiwan has become the second country who had commenced construction of ABWR (Lungmen nuclear power plant (NPP)). It has two identical units with 3,926 MWt rated thermal power each and 52.2*106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637*106 kg/hr at rated power. There are 10 RIPs in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 1,450 rpm. In this paper, the steam-line break inside containment LOCA transient data from FSAR is used to verify and establish the Lungmen TRACE model. It compares those important thermal parameters at steady state and transient, such as the dome pressure of reactor vessel, steam flow, feedwater flow, and core flow, etc.. It was concluded that the results of TRACE calculations are in agreement with those from FSAR. In summary, our studies concluded that the analysis results trends of Lungmen NPP TRACE model are roughly consistence with FSAR data for the steam-line break inside

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

    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

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

    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.

  9. Completion of latest ABWR 'SHIKA Unit 2' construction

    The Shika Nuclear Power Station Unit No.2 of the Hokuriku Electric Power Company, Inc. is the first Advanced BWR unit built in Japan by a single contractor and it is among the largest nuclear power stations in Japan. Its construction started in August 1999 when the first construction permit was issued. The design and construction of the plant was carried out with utmost care for betterment of operational safety, reliability and economy. The construction advanced on schedule and the plant entered its commercial operation in March 2006 as planned. Hitachi, Ltd. supplied the entire plant from design, fabrication and construction including the reactor and steam turbine generation system. In the design and construction of the plant, the most advanced technology has been applied in order to match the civil construction process and aim to supply safest, reliable and economical power plant. (author)

  10. ABWR (K-6/7) construction experience (computer-based safety system)

    TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

  11. Rewetting an auto-wave change of boiling modes

    One studied heat exchange under auto-wave change of boiling metastable mode over to the stable one. One introduced a classification of boiling curves depending on temperature gradient, flow rate directions and temperature wave motion rate. One advanced the hypothesis according to which the essential change of heat exchange patterns within temperature wave range was caused by boiling local nonequilibrium state. One investigated experimentally heat exchange under water boiling in pipes under auto-wave change of boiling conditions within 3-10 MPa pressure range within wide range of steam contents and mass rates. Paper discusses application possibility of the derived data when simulating emergency conditions in reactor core

  12. MELCOR 1.8.2 calculations of selected sequences for the ABWR

    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

  13. Aspects of subcooled boiling

    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.

  14. Nucleate boiling heat transfer

    Saiz Jabardo, J.M. [Universidade da Coruna (Spain). Escola Politecnica Superior], e-mail: mjabardo@cdf.udc.es

    2009-07-01

    Nucleate boiling heat transfer has been intensely studied during the last 70 years. However boiling remains a science to be understood and equated. In other words, using the definition given by Boulding, it is an 'insecure science'. It would be pretentious of the part of the author to explore all the nuances that the title of the paper suggests in a single conference paper. Instead the paper will focus on one interesting aspect such as the effect of the surface microstructure on nucleate boiling heat transfer. A summary of a chronological literature survey is done followed by an analysis of the results of an experimental investigation of boiling on tubes of different materials and surface roughness. The effect of the surface roughness is performed through data from the boiling of refrigerants R-134a and R-123, medium and low pressure refrigerants, respectively. In order to investigate the extent to which the surface roughness affects boiling heat transfer, very rough surfaces (4.6 {mu}m and 10.5 {mu}m ) have been tested. Though most of the data confirm previous literature trends, the very rough surfaces present a peculiar behaviour with respect to that of the smoother surfaces (Ra<3.0 {mu}m). (author)

  15. Prediction of a subcooled boiling flow with mechanistic wall boiling and bubble size models

    Subcooled boiling is one of the crucial phenomena for the design, operation and safety analysis of a nuclear power plant. In recent years, developers of multiphase CFD (Computational Fluid Dynamics) codes focused their development activity on the mechanistic prediction of DNB (Departure from Nucleate Boiling) in PWR. Wall boiling model is one of the key parameters for this purpose. In order to enhance prediction capability of the subcooled boiling flow, an advanced wall boiling model consisting of a mechanistic bubble departure model (Klausner et al., 1993), Hibiki et al.'s (2009) active nucleate site model and Cole's bubble departure frequency model was explored for the CFD code. To ensure a wide range applicability of the advanced wall boiling model, each model was evaluated separately according to the flow conditions such as pressure, temperature and flow rate. Finally, the advanced wall boiling model was implemented into the STAR-CD as a form of user FORTRAN file. One of the other important parameters for an accurate prediction of the subcooled boiling flow is bubble size which governs interfacial transfer terms between two phases. In this study, the S-gamma model, which was developed for the STAR-CD (Lo, 2006), was applied as a bubble size model. For the validation of the present wall boiling and bubble size models, benchmark calculations were carried out against SUBO and DEBORA subcooled boiling flow data. Working fluid of SUBO test is steam/water and its pressure condition is about 2 bars. In contrast to this, working fluid of DEBORA test is Refrigerant-12 (R-12) and phasic density ratio of the tests is equivalent to that of steam/water around 90 to 170 bars. Therefore, present benchmark calculation covers wide range pressure condition of steam/water. The calculation results confirms that the new mechanistic wall boiling and bubble size models follow well the tendency on the change of flow conditions and they can be applicable to the wide range of flow

  16. Radiolysis of boiling water

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  17. The ultimate response guideline simulation and analysis by using (TRACE) for Lungmen ABWR nuclear power plant

    In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

  18. The ultimate response guideline simulation and analysis by using (TRACE) for Lungmen ABWR nuclear power plant

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

    2015-07-15

    In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

  19. 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

    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

  20. Two approaches to meeting the economic challenge for advanced BWR designs

    This paper presents the design overview and approach to addressing the aforementioned economic challenges for two Advanced BWR designs. The first plant is the ABWR and the second is the ESBWR. The ABWR relies on proven technology and components and an extensive infrastructure that has been built up over the last 20 years. Because it has proven and standards safety systems it has very limited uncertainty regarding licensing. Finally, it relies on the economies of scale and overall design flexibility to improve the overall economics of power generation. The ESBWR on the other hand has taken an innovative approach to reduce systems and components to simplify the overall plant to improve plant economics. The overall plant design is indeed simpler, but improved economics required reliance on some economies of scale also. This design embodied in the ESBWR, also has minimized the overall development cost by utilizing features and components from the ABWR and SBWR technology programs

  1. Spectral analysis of boiling sound

    Experimental apparatus, measurements and spectral analysis of boiling sound are described as observed in subcooled boiling of water on a Pt-wire. The results indicate the existence of a strong relation between the intensity and the average frequency of the boiling sound vs heat flux. (author)

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

    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)

  3. RETRAN analysis results of feedwater pump trip transient for Lungmen ABWR Plant

    Highlights: → The RETRAN model was used to predict one feedwater pump trip (FWPT) transient. → The result shows that the margin sustains at least 30 cm above the L3 setpoint. → The unavailable motor driven pump case eventually actuates the low level scram signal. → The lowest load line case without motor driven pump still actuates the L3 scram. - Abstract: The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient. The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low

  4. Geysering in boiling channels

    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.

  5. Natural Circulation with Boiling

    A number of parameters with dominant influence on the power level at hydrodynamic instability in natural circulation, two-phase flow, have been studied experimentally. The geometrical dependent quantities were: the system driving head, the boiling channel and riser dimensions, the single-phase as well as the two phase flow restrictions. The parameters influencing the liquid properties were the system pressure and the test section inlet subcooling. The threshold of instability was determined by plotting the noise characteristics in the mass flow records against power. The flow responses to artificially obtained power disturbances at instability conditions were also measured in order to study the nature of hydrodynamic instability. The results presented give a review over relatively wide ranges of the main parameters, mainly concerning the coolant performance in both single and parallel boiling channel flow. With regard to the power limits the experimental results verified that the single boiling channel performance was intimately related to that of the parallel channels. In the latter case the additional inter-channel factors with attenuating effects were studied. Some optimum values of the parameters were observed

  6. Start of the construction of initial advanced BWRs

    The construction of first advanced BWRs (ABWR), No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., was started, and the commercial operation is scheduled to begin in 1996 and 1997. The ABWR with the unit capacity of 1356 MW which is the largest class in the world has been developed aiming at the further improvement of safety, operation performance, economical efficiency and so on. Hitachi Limited carried out the detailed review on new parts design and expected the perfection so as to make the first machine into the plant maintaining high reliability. The course of development of the ABWR, the outline of No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, the review of the design of new parts, the establishment of the technology of manufacturing RCCVs, the verification of the maintenance and checkup of internal pumps and improved type control rod driving mechanism and so on are reported. The ABWR is Japanese third improved type standard plant, but also it is expected to be adopted worldwide. (K.I.)

  7. Revision of nucleated boiling mechanisms

    The boiling occurrence plays an important role in the power reactors energy transfer. But still, there is not a final theory on the boiling mechanisms. This paper presents a critical analysis of the most important nucleated boiling models that appear in literature. The conflicting points are identified and experiments are proposed to clear them up. Some of these experiments have been performed at the Thermohydraulics laboratory (Bariloche Atomic Center). (Author)

  8. Dispersed flow film boiling

    Dispersed flow film boiling is the heat transfer regime that occurs at high void fractions in a heated channel. The way this transfer mode is modelled in the NRC computer codes (RELAP5 and TRAC) and the validity of the assumption and empirical correlations used is discussed. An extensive review of the theoretical and experimental work related with heat transfer to highly dispersed mixtures reveals the basic deficiencies of these models: the investigation refers mostly to the typical conditions of low rate bottom reflooding, since the simulation of this physical situation by the computer codes has often showed poor results. The alternative models that are available in the literature are reviewed, and their merits and limits are highlighted. The modification that could improve the physics of the models implemented in the codes are identified. (author) 13 figs., 123 refs

  9. Steady State Vapor Bubble in Pool Boiling

    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.

  10. Steady State Vapor Bubble in Pool Boiling.

    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

  11. Bandages of boiled potato peels.

    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

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

    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)

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

    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)

  14. 3RIP trip startup test simulation of TRACE/PARCS model for Lungmen ABWR under different power and flow conditions

    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.

  15. Sodium hydroxide injection passivation work for the reactor water clean-up system in a new ABWR plant

    Several studies have identified that Co-58 and Co-60 as the primary source of radiation build up on out-of-core components in new BWR plants. The deposition rate of Co on stainless steel and carbon steel is shown to be controlled mainly by the thickness of oxide films and its morphology formed through pretreatment. The passivation treatment was implemented accordingly at Lungmen unit 1 in an ABWR plant in September 2010. It is determined that the passivation conditions should be maintained at the temperature of 180∼230 deg. C, pH of 8.0∼8.5 and dissolved oxygen content over 400 ppb. The films would provide effective protection against radioactive deposition. The application of the pre-filming process on piping before the pre-operation is done during the flow induced vibration test (FIV) period. The protectiveness of stable magnetite can be increased by the pH control under the specific condition. The pre-filming control process and evaluation of passivation effectiveness is discussed in detail based on the surface analysis of the passivated specimens. Many efforts have been devoted to sodium hydroxide injection method for pH control of the system through the filter demineralizer under smooth operation. A comparison of test specimens on the properties of oxide film formed between laboratory and in-plant tests through alkaline treatment are also shown in this report. (authors)

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

    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. Surface boiling of superheated liquid

    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

  18. Surface boiling of superheated liquid

    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.

  19. Introduction of characteristics and construction methods for Lungmen NPP

    Daewoo Engineering and Construction Co., Ltd. has taken part in the construction of ABWR (Advance Boiling Water Reactor of capacity of 1,350MW) at Lungmen site in Taiwan since 1999 as a member of Joint Company for the Nuclear Island Civil Work and Nuclear Island Equipment and Piping Work. This paper discusses the introduction of NSSS of ABWR and the major construction methods applied to Lungmen Nuclear Power Plants. Even though there is no BWR or ABWR type of Nuclear Power Plants built in Korea, the information and data related to the newly advanced construction technologies which has been already applied successfully is very important for the future overseas expansion

  20. LMFBR safety and sodium boiling

    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.

  1. Molecular geometry and boiling related thermodynamic properties

    Highlights: ► Molecular geometric factors were found to be important determinants for boiling entropy and thus the boiling temperature. ► Only four molecular geometric factors were used in the study. ► A group contribution method was used to calculate enthalpy of boiling. ► The proposed method is simple and the estimations are in good agreement with experimental values. - Abstract: Boiling related thermodynamic properties are important parameters in research. In this study, a model integrating both additive groups and non-additive molecular geometric factors has been developed for the calculation of boiling enthalpy, entropy and temperature. The calculated values are in good agreement with the measured values of 470 compounds. This model provides a simple and accurate estimation of enthalpy of boiling, entropy of boiling and boiling temperatures with absolute average errors of 0.62 kJ/mol, 1.15 J/K · mol and 7.13 K respectively.

  2. Numerical Modeling and Investigation of Boiling Phenomena

    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...

  3. The nucleate pool boiling dilemma

    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.)

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

    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

  5. Advanced nuclear plants meet the economic challenge

    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.)

  6. Thermosyphon boiling in vertical channels

    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.

  7. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  8. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    Tuerkcan, E. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kozma, R. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Nabeshima, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1993-01-01

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  9. Duality of boiling systems and uncertainty phenomena

    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.

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

    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...

  11. Heat transfer research on enhanced heating surfaces in flow boiling in a minichannel and pool boiling

    Graphical abstract: - Highlights: • Application of enhanced surfaces in boiling heat transfer. • Flow and pool boiling heat transfer on the heating surfaces with mini-recesses. • Minichannel (horizontal) with the enhanced heating wall. • Determination of heat transfer coefficients and boiling curves. • Comparative experimental data analysis for flow and pool boiling heat transfer. - Abstract: The paper focuses on the analysis of the enhanced surfaces in such applications as boiling heat transfer. The surfaces have similar geometric parameters for the surface development. Two testing measurement modules with enhanced heating surfaces are used independently, one for flow boiling and the other – for pool boiling research. The heating surfaces with mini-recesses which contact boiling liquid are made by spark erosion. Flow boiling is studied when FC-72 flows through a horizontally positioned minichannel and its bottom wall is heated. These experiments were carried out during under a pressure slightly higher than the atmospheric one. Pool boiling experiments were conducted with FC-72 at atmospheric pressure in the vessel using enhanced sample as the bottom heating surface. Comparison of results for flow and pool boiling indicates that obtained heat transfer coefficients are a few times higher for pool boiling in the boiling incipience conditions. There are basic differences in the local heat transfer coefficients during the development of flow boiling in a minichannel, depending on the location along the flow in the channel. In the subcooled boiling area, heat transfer coefficients are low. In developed boiling, they are high, but they decrease when the amount of vapour in the liquid–vapour mixture rises

  12. Subcooled flow boiling experiments and numerical simulation for a virtual reactor development

    Subcooled flow boiling experiments and numerical simulations using a Lattice Boltzmann model will be performed at City College of New York as part of the DOE Nuclear HUB project, Consortium for Advanced Simulation of Light Water Reactors (CASL). The experiments being performed include pool boiling from a platinum wire, subcooled flow boiling in a vertical tube and single air bubble injection into a turbulent water stream. Preliminary experiments have been performed to measure the bubble size, shape and motion in an adiabatic experiment involving air bubble injection into water flowing in a vertical annulus, as well as PIV measurements of liquid flow field in a subcooled flow boiling experiment. An advanced thermal Finite Element Lattice Boltzmann Model is being developed to predict the pool and flow boiling experiments. After the validation of the code, improved constitutive relations for subcooled flow boiling will be developed for use in 3-D CFD models. The present work is expected to contribute to the development of a multi-scale, multi-physics model of a PWR in the CASL project. (author)

  13. High Pressure Boiling Water Reactor

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  14. Boiling flow through diverging microchannel

    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.

  15. Signal processing for boiling noise detection

    The present paper deals with investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. (author). 6 refs, figs

  16. Transition boiling heat transfer during reflooding transients

    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

  17. Boiling water reactor simulator. Workshop material

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

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

    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

  19. 21 CFR 872.6710 - Boiling water sterilizer.

    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...

  20. Using Boiling for Treating Waste Activated Sludge

    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.

  1. How To Boil the Perfect Egg

    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

  2. Boiling of subcooled water in forced convection

    As a part of a research about water cooled high magnetic field coils, an experimental study of heat transfer and pressure drop is made with the following conditions: local boiling in tubes of small diameters (2 and 4 mm), high heat fluxes (about 1000 W/cm2), high coolant velocities (up to 25 meters/s), low outlet absolute pressures (below a few atmospheres). Wall temperatures are determined with a good accuracy, because very thin tubes are used and heat losses are prevented. Two regimes of boiling are observed: the establishment regime and the established boiling regime and the inception of each regime is correlated. Important delays on boiling inception are also observed. The pressure drop is measured; provided the axial temperature distribution of the fluid and the axial distributions of the wall temperatures, in other words the axial distribution of the heat transfer coefficients under boiling and non boiling conditions, at the same heat flux or the same wall temperatures, are taken in account, then total pressure drop can be correlated, but probably under certain limits of void fraction only. Using the same parameters, it seems possible to correlate the experimental values on critical heat flux obtained previously, which show very important effect of length and hydraulic diameter of the test sections. (authors)

  3. Experimental study on transient boiling heat transfer

    Boiling phenomena can be found in the everyday life, thus a lot of studies are devoted to them, especially in steady state conditions. Transient boiling is less known but still interesting as it is involved in the nuclear safety prevention. In this context, the present work was supported by the French Institute of Nuclear Safety (IRSN). In fact, the IRSN wanted to clarify what happens during a Reactivity-initiated Accident (RIA). This accident occurs when the bars that control the nuclear reactions break down and a high power peak is passed from the nuclear fuel bar to the surrounding fluid. The temperature of the nuclear fuel bar wall increases and the fluid vaporises instantaneously. Previous studies on a fuel bar or on a metal tube heated by Joule effect were done in the past in order to understand the rapid boiling phenomena during a RIA. However, the measurements were not really accurate because the measurement techniques were not able to follow rapid phenomena. The main goal of this work was to create an experimental facility able to simulate the RIA boiling conditions but at small scale in order to better understand the boiling characteristics when the heated-wall temperature increases rapidly. Moreover, the experimental set-up was meant to be able to produce less-rapid transients as well, in order to give information on transient boiling in general. The facility was built at the Fluid-Mechanics Institute of Toulouse. The core consists of a metal half-cylinder heated by Joule effect, placed in a half-annulus section. The inner half cylinder is made of a 50 microns thick stainless steel foil. Its diameter is 8 mm, and its length 200 mm. The outer part is a 34 mm internal diameter glass half cylinder. The semi-annular section is filled with a coolant, named HFE7000. The configuration allows to work in similarity conditions. The heated part can be place inside a loop in order to study the flow effect. The fluid temperature influence is taken into account as

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

    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...

  5. On the frontier of boiling curve and beyond design of its origin

    An advanced approach of Extended Design of the Boiling Curve beyond its origin is proposed. It is developed from the fact that both CHF (Critical Heat Flux) and rewetting affect the Boiling Curve on the heating surface through two simultaneous processes taking place on both sides of the heating surface. The first is two-phase flow thermal-hydraulics with resultant heat transferred from the heating surface to the coolant. The second one is the heat conduction through material itself, allied with the balance of generated and accumulated energy. Both of these processes are triggered by the change in HTC (Heat Transfer Coefficient) on the heating surface, which accordingly influences the Boiling Curve. Depending on direction of the Transition - from nucleate to film boiling or vice versa - these processes act differently and direct the Boiling Curve to diverse paths. The proposed physically based concept recognises this fact and introduces HTC as the triggering parameter with instant effect. It is implemented in the subchannel code COBRA 3-CP providing stable rewetting which has been deficient in COBRA since its origin. Results of validation and obtained agreements with transient measured data prove legality of the advanced concept of Boiling Curve. This approach is being used for transient analyses of PWR (Pressurised Water Reactor) gaining benefits from properly predicting the rewetting. The method is well-qualified to be applied also in other thermal-hydraulic codes like COBRA/TRAC, COBRA-TF, TRAC and/or RELAP, where the classical steady-state and poolboiling approach has been originally implemented. (author)

  6. How does surface wettability influence nucleate boiling?

    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).

  7. Film boiling heat transfer during reflood process

    From Westinghouse's Full Length Emergency Cooling Heat Transfer (FLECHT) test data and the previous studies on the film boiling, local subcooling is found to be a dominant factor in the film boiling heat transfer, existing in the reflood process. By experiment, the correlation was obtained between saturated film boiling heat transfer coefficient h sub(c), sat and subcooled h sub(c), sub. The h sub(c), sat is similar to Bromley's expression, but the value differs from his. The ratio of h sub(c') sub to h sub(c') sat is expressed with the local coolant subcooling T sub(sub) (0C) as h sub(c') sub/h sub(c') sat = 1 + 0.025 ΔT sub(sub). The results in experiment are predicted by this formula with error +- 20%. (auth.)

  8. Thermodynamics of Flow Boiling Heat Transfer

    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.

  9. Relevant thermal hydraulic aspects of advanced reactors design: status report

    This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs

  10. Nucleate pool-boiling heat transfer - I. Review of parametric effects of boiling surface

    The objective of this paper is to assess the state-of-the-art of heat transfer in nucleate pool-boiling. Therefore, the paper consists of two parts: part I reviews and examines the effects of major boiling surface parameters affecting nucleate-boiling heat transfer, and part II reviews and examines the existing prediction methods to calculate the nucleate pool-boiling heat transfer coefficient (HTC). A literature review of the parametric trends points out that the major parameters affecting the HTC under nucleate pool-boiling conditions are heat flux, saturation pressure, and thermophysical properties of a working fluid. Therefore, these effects on the HTC under nucleate pool-boiling conditions have been the most investigated and are quite well established. On the other hand, the effects of surface characteristics such as thermophysical properties of the material, dimensions, thickness, surface finish, microstructure, etc., still cannot be quantified, and further investigations are needed. Particular attention has to be paid to the characteristics of boiling surfaces. (author)

  11. Evaluation of heat transfer at dispersed film boiling region and reflood blockage by downcomer boiling

    The large nuclear power plant like APR1400 have a emergency core cooling system (ECCS) for large break loss of coolant accident (LBLOCA). To evaluate the cooling capacity of ECCS, it is important to analysis the heat transfer at dispersed film boiling region and to evaluate the amount of reflood. During the reflood, boiling occurs at the downcomer of vessel and the boiling play a role of blockage and hence the amount of inflow is reduced. Therefore, the phenomena also will be evaluated. This study is composed of three key subjects. One is the study about the heat transfer at dispersed film boiling. In this subject, the final goal is to develop a boiling model. For this, we will analysis the experimental results and other correlation. and the mechanisms will be also compared with each other. The new model will be developed including recent experimental results. Second one is to do a experimental works about the amount of inflow under the downcomer boiling simulation. In this experiment, the air-water is working fluid. the bubble dynamics, pressure drop of two phase flow, blockage effect and etc. will be observed. and some of these will be quantified. Third subjects is to control the boiling heat transfer coefficient. Here, the method will be various surface treatment.

  12. The boiling point of stratospheric aerosols.

    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.

  13. Water boiling kinetic in rapid decompression

    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

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

    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)

  15. Fuel assembly for a boiling water reactor

    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.)

  16. Heat transfer coefficient for boiling carbon dioxide

    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...

  17. Heat transfer coeffcient for boiling carbon dioxide

    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...

  18. Classic and Hard-Boiled Detective Fiction.

    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…

  19. A dry-spot model of critical heat flux applicable to both pool boiling and subcooled forced convection boiling

    A study has been performed to predict CHF in pool boiling and subcooled forced convection boiling using the dry-spot model presented by the authors and existing correlations for heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling. Comparisons of the model predictions with experimental data for pool boiling of water and subcooled upward forced convection boiling of water in vertical, uniformly-heated round tubes have been performed and the parametric trends of CHF have been investigated. The results of the present study strongly support the validity of physical feature of the present model on the CHF mechanism in pool boiling and subcooled forced convection boiling. To improve the prediction capability of the present model, further works on active site density, bubble departure diameter and suppression factor in subcooled boiling are needed

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

    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

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

    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.

  2. Boiling Heat Transfer on Superhydrophilic, Superhydrophobic, and Superbiphilic Surfaces

    Betz, Amy Rachel; Kim, Chang-Jin 'CJ'; Attinger, Daniel

    2012-01-01

    With recent advances in micro- and nanofabrication, superhydrophilic and superhydrophobic surfaces have been developed. The statics and dynamics of fluids on these surfaces have been well characterized. However, few investigations have been made into the potential of these surfaces to control and enhance other transport phenomena. In this article, we characterize pool boiling on surfaces with wettabilities varied from superhydrophobic to superhydrophilic, and provide nucleation measurements. The most interesting result of our measurements is that the largest heat transfer coefficients are reached not on surfaces with spatially uniform wettability, but on biphilic surfaces, which juxtapose hydrophilic and hydrophobic regions. We develop an analytical model that describes how biphilic surfaces effectively manage the vapor and liquid transport, delaying critical heat flux and maximizing the heat transfer coefficient. Finally, we manufacture and test the first superbiphilic surfaces (juxtaposing superhydrophobic ...

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

    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

  4. Numerical simulation of subcooled flow boiling

    Park, Won Cheol

    Sub-cooled flow boiling in a U-bend has been examined using numerical methods. An Eulerian/Eulerian mathematical description was used with a multiphase computational algorithm to predict several types of flows and to examine sub-cooled flow boiling. As a prelude to the study of sub-cooled boiling and two-phase flows, single-phase laminar and turbulent flows in a U-bend were investigated. Air-water bubbly up flow in a vertical straight duct followed by a U-bend with heat transfer was analyzed. In such a flow, as the flow develops through the U-bend the bubbles move from center and outer wall toward inner wall. After half way through the U-bend, the fluids do not have sufficient time for complete reorganization in the presence of centrifugal forces and the pressure gradients. After the U-bend, the bubbles finally reach the original distribution in about forty diameters. The heat transfer in the U-bend was also calculated and as expected heat transfer rate on the outer wall is higher than on the inner wall. For air-water bubbly two-phase flow, Nusselt numbers in the U-bend can be as high as 400 percent of the value in the straight duct on one of the walls. The method of partitioned wall heat flux was used to study sub-cooled flow boiling. For sub-cooled flow boiling in a U-bend, axial and lateral velocity distributions as well as quality and void fraction variations were analyzed. Computed axial and lateral variations of void fraction compare favorably with existing experimental data. As expected, the pressure drop for bubbly flow through the U-bend is larger than for single-phase flow by as much as fifty percent. Computed pressure drop for flow with phase change falls between the predictions of two different correlations in the literature, and thus seems reasonable. Predictions of heat transfer and void fraction under sub-cooled flow boiling using two-fluid models need better quantitative knowledge related to the mechanisms associated with bubble growth and

  5. Transition from natural convection or nucleate boiling regime to nucleate boiling or film boiling regime caused by a rapid pressure reduction in highly pressurized and subcooled water

    Transient boiling processes caused by exponentially decreasing system pressures with various decreasing pressure-reduction periods from the initial heat flux on a horizontal cylinder in a pool of highly subcooled water measured were divided into three groups for low and intermediate initial heat flux in natural convection regime and for high initial heat flux in nucleate boiling. The transitions from low initial heat flux values to stable nucleate boiling occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values to stable film boiling occurred for the short pressure-reduction period values, although those to stable nucleate boiling occurred for the long pressure-reduction period values. The mechanism of transient boiling process caused by an exponentially decreasing system pressure with a decreasing pressure-reduction period from an initial heat flux on a horizontal cylinder in a pool of highly subcooled water was clarified on the graph of α/q0.7 versus system pressure with the curves of corresponding fully developed nucleate boiling, incipient nucleate boiling due to unflooded cavities with vapor, and incipient heterogeneous spontaneous nucleation (HSN) due to flooded cavities without vapor. The transitions to stable nucleate boiling from the low initial heat flux values occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values in natural convection and nucleate boiling to stable film boiling occurred due to the HSN for short pressure-reduction period values; however those to stable nucleate boiling occurred for long pressure-reduction values. (author)

  6. Dynamic Bubble Behaviour during Microscale Subcooled Boiling

    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.

  7. Boiling Heat Transfer in Circulating Fluidized Beds

    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.

  8. Boiling Heat Transfer in Circulating Fluidized Beds

    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.

  9. Enzyme engineering reaches the boiling point

    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...

  10. Self-propelled film-boiling liquids

    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.

  11. Nucleate boiling of oxygen under reduced gravity

    Kirichenko, Y.A.; Gladchenko, G.M.; Rusanov, K.V.

    1986-03-01

    Experimental results are presented on the coefficients of nucleate boiling heat transfer of oxygen under the conditions of low loading factors at elevated pressures. Based on the statistical distribution of the separation bubble radii and distances between the nucleation sites, a relation is obtained which provides a satisfactory description of the function ..cap alpha..(q) in case of deteriorated heat transfer at eta = g/g/sub n/ < 0.1.

  12. Nucleate boiling of oxygen under reduced gravity

    Experimental results are presented on the coefficients of nucleate boiling heat transfer of oxygen under the conditions of low loading factors at elevated pressures. Based on the statistical distribution of the separation bubble radii and distances between the nucleation sites, a relation is obtained which provides a satisfactory description of the function α(q) in case of deteriorated heat transfer at eta = g/g/sub n/ < 0.1

  13. Serious accidents on boiling water reactors (BWR)

    This short document describes, first, the specificities of boiling water reactors (BWRs) with respect to PWRs in front of the progress of a serious accident, and then, the strategies of accident management: restoration of core cooling, water injection, core flooding, management of hydrogen release, depressurization of the primary coolant circuit, containment spraying, controlled venting, external vessel cooling, erosion of the lower foundation raft by the corium). (J.S.)

  14. European simplified boiling water reactor (ESBWR) plant

    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

  15. Flow boiling test of GDP replacement coolants

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C4F10 and C4F8, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C4F10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C4F10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  16. Evaluation of onset of nucleate boiling models

    This article discusses available models and correlations for predicting the required heat flux or wall superheat for the Onset of Nucleate Boiling (ONB) on plain surfaces. It reviews ONB data in the open literature and discusses the continuing efforts of Heat Transfer Research, Inc. in this area. Our ONB database contains ten individual sources for ten test fluids and a wide range of operating conditions for different geometries, e.g., tube side and shell side flow boiling and falling film evaporation. The article also evaluates literature models and correlations based on the data: no single model in the open literature predicts all data well. The prediction uncertainty is especially higher in vacuum conditions. Surface roughness is another critical criterion in determining which model should be used. However, most models do not directly account for surface roughness, and most investigators do not provide surface roughness information in their published findings. Additional experimental research is needed to improve confidence in predicting the required wall superheats for nucleation boiling for engineering design purposes. (author)

  17. CFD simulation of DEBORA boiling experiments

    Rzehak, Roland; Krepper, Eckhard

    2012-08-01

    In this work we investigate the present capabilities of computational fluid dynamics for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. This kind of modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant non-dimensional 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, bubble size and liquid temperature as well as axial profiles of wall temperature. After reviewing the theoretical and experimental basis of correlations used in the ANSYS CFX model used for the calculations, we give a careful assessment of the necessary recalibrations to describe the DEBORA tests. The basic CFX model is validated by a detailed comparison to the experimental data for two selected test cases. Simulations with a single set of calibrated parameters are found to give reasonable quantitative agreement with the data for several tests within a certain range of conditions and reproduce the observed tendencies correctly. Several model refinements are then presented each of which is designed to improve one of the remaining deviations between simulation and measurements. Specifically we consider a homogeneous MUSIG model for the bubble size, modified bubble forces, a wall function for turbulent boiling flow and a partial slip boundary condition for the liquid phase. Finally, needs for further model developments are identified and promising directions discussed.

  18. CFD for Subcooled Flow Boiling: Parametric Variations

    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.

  19. Study on film boiling heat transfer and minimum heat flux condition for subcooled boiling, 1

    In the present paper, film boiling heat transfer and minimum heat flux condition were experimentally studied for subcooled pool boiling of water at atmospheric pressure from a platinum sphere (D = 10 mm). Transient tests of subcooled boiling were conducted from an initial temperature of the sphere about 1500 K. Experimental parameters were liquid subcooling (0 ∼ 75 K) and the depth of immersion of the sphere (0.75 D ∼ 3.0 D). The obtained boiling curves indicated that the ideal depth was 1.8 D. Even for large subcooling conditions, the measured temperature at the minimum heat flux point for such a depth did not remarkably exceed the maximum superheat of water. Further, accounting for the density-viscosity ratio, the analytical equation for subcooled film boiling derived by Hamill and Baumeister was modified. This modified equation was in good agreement with experimental data of water and freon-11 under various conditions of surface geometry, size and system pressure. (author)

  20. Assessment of Boiling Model in a Computational Analysis of the Subcooled Boiling Flow

    Two-phase flow phenomena are known to be crucial for a nuclear reactor safety, such as a subcooled boiling at the downcomer during a Large-Break Loss-of- Coolant Accident (LBLOCA). For the analysis of a two-phase flow, the two-fluid model is considered as a state-of-the-art model which deals with the mass, momentum and energy of each phase. The interfacial area concentration (IAC), which is defined as the area of interface per unit mixture volume, is one of the most significant parameters in the two-fluid model. In order to resolve the problems of the conventional models for an IAC, an interfacial area transport equation has been developed for an adiabatic bubbly flow or nucleate boiling flow. For the investigation of a boiling flow with a dynamic modeling of the interfacial structure, this study focuses on the development of a computational fluid dynamics (CFD) code with implementing an interfacial area transport equation. A benchmark problem for the subcooled boiling flow is analyzed with the developed code so that the sensitivity on the boiling model can be analyzed

  1. Stability monitoring for boiling water reactors

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  2. Battery thermal management by boiling heat-transfer

    Highlights: • A thermal management scheme based on boiling heat-transfer is investigated. • Cooling capacity of the working fluid compared to that of air is investigated. • Battery gets fluid to boil, thus boiling heat-transfer occurs from battery to fluid. • Boiling process thermally homogenises the battery. • Boiling process can be influenced by pressure variation. - Abstract: In this study, the ability of a boiling process to thermally condition (homogenisation and cooling) batteries is investigated. Thereto, a series of experiments are performed and discussed. Subjects that are treated are the dielectric property of the proposed cooling fluid, its cooling capability compared to that of air, the ability of the boiling fluid to thermally homogenise a battery and the influence of pressure variation on the boiling process. It turns out that the proposed cooling fluid conducts no electricity, has good cooling characteristics compared to those of air and, when boiling, is able to thermally homogenise the battery. Furthermore, pressure variation seems to offer a good method to regulate the boiling process

  3. Boiling radial flow in fractures of varying wall porosity

    Barnitt, Robb Allan

    2000-06-01

    The focus of this report is the coupling of conductive heat transfer and boiling convective heat transfer, with boiling flow in a rock fracture. A series of experiments observed differences in boiling regimes and behavior, and attempted to quantify a boiling convection coefficient. The experimental study involved boiling radial flow in a simulated fracture, bounded by a variety of materials. Nonporous and impermeable aluminum, highly porous and permeable Berea sandstone, and minimally porous and permeable graywacke from The Geysers geothermal field. On nonporous surfaces, the heat flux was not strongly coupled to injection rate into the fracture. However, for porous surfaces, heat flux, and associated values of excess temperature and a boiling convection coefficient exhibited variation with injection rate. Nucleation was shown to occur not upon the visible surface of porous materials, but a distance below the surface, within the matrix. The depth of boiling was a function of injection rate, thermal power supplied to the fracture, and the porosity and permeability of the rock. Although matrix boiling beyond fracture wall may apply only to a finite radius around the point of injection, higher values of heat flux and a boiling convection coefficient may be realized with boiling in a porous, rather than nonporous surface bounded fracture.

  4. Flow boiling heat transfer at low liquid Reynolds number

    Full text of publication follows: In view of the significance of a heat transfer correlation of flow boiling at conditions of low liquid Reynolds number or liquid laminar flow, and very few existing correlations in principle suitable for such flow conditions, this study is aiming at developing a heat transfer correlation of flow boiling at low liquid Reynolds number conditions. The obtained results are as follows: 1. A new heat transfer correlation has been developed for saturated flow boiling at low liquid Reynolds number conditions based on superimposition of two boiling mechanisms, namely convective boiling and nucleate boiling. In the new correlation, two terms corresponding to the mechanisms of nucleate boiling and convective boiling are obtained from the pool boiling correlation by Forster and Zuber and the analytical annular flow model by Hewitt and Hall-Taylor, respectively. 2. An extensive database was collected for saturated flow boiling heat transfer at low liquid Reynolds number conditions, including data for different channels geometries (circular and rectangular), flow orientations (vertical and horizontal), and working fluids (water, R11, R12, R113). 3. An extensive comparison of the new correlation with the collected database shows that the new correlation works satisfactorily with the mean deviation of 16.6% for saturated flow boiling at low liquid Reynolds number conditions. 4. The detailed discussion reveals the similarity of the newly developed correlation for flow boiling at low liquid Reynolds number to the Chen correlation for flow boiling at high liquid Reynolds number. The Reynolds number factor F can be analytically deduced in this study. (authors)

  5. Subcooled boiling model to simulate upward vertical flow boiling at low pressures

    A new model for upward vertical subcooled flow boiling at low pressure is proposed. The model considers the most relevant closure relationships of one-dimensional thermalhydraulic codes that are important for prediction of vapor contents in the channel: wall evaporation model, condensation model, flow regime transition criterion and drift-flux model. The new model was incorporated in the current version of the thermal-hydraulic computer code RELAP5/MOD3.2.2 Gamma. The modified code was validated against a number of published low-pressure subcooled boiling experiments, and in contrast to the current code, shows good agreement with experimental data. The presented analysis also leads to a better understanding of the basic mechanisms of subcooled flow boiling at low pressure.(author)

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

    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

  7. Regression analysis of post-CHF flow boiling data

    The successful application of statistical analysis in systematic investigations of heat transfer data for boiling water beyond the critical heat flux is described. Multiple linear regression analysis together with statistical tests of correlations and data were used in this study. Data from a number of experiments encompassing film and transition boiling in several geometries were correlated by boiling regime, by geometry, and in aggregate. Error estimates and uncertainty bounds were specified for all such correlations. (U.S.)

  8. Boiling of an Emulsion in a Yield Stress Fluid

    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...

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

    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...

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

    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...

  11. Boiled Water Temperature Measurement System Using PIC Microcontroller

    A.T.KARUPPIAH, AZHA. PERIASAMY, P.RAJKUMAR

    2013-01-01

    The measurement system for temperature of boiled water is a critical task in industry. In this paper we designed and implemented a PIC micro controller based boiled water temperature measurement system using PIC 18F452 and national semiconductors LM35 temperature sensor. The designing system is used to measure the tank I boiled water temperature value. If the temperature value reaches the set value high temperature relay board becomes ON to control the solenoid valve. The high temperature of ...

  12. Downflow film boiling in a rod bundle at low pressure

    A series of low pressure downflow film boiling heat transfer experiments were conducted in a 14-foot (4.27 m) long electrically heater rod bundle containing 336 heater rods. The resulting data was compared with the Dougall-Rohsenow dispersed flow film boiling correlation. The data was found to lie below this correlation as the quality was increased. It is believed that buoyancy effects decreased the heat transfer in downflow film boiling. (author)

  13. Mass exchange calculation in a wall layer when water boiling

    Physical sense and mass exchange characteristics of liquid near-the-wall layer under boiling conditions were attempted to be stated. Equations of material and thermal balance were used to describe the mass exchange characteristics. Technique to calculate circulation ratio in the near-the-wall layer under boiling of under-heated and saturated water was suggested on the basis of the derived expressions. Comparison results of calculated and experimental data were analyzed for full-scale boiling

  14. Boiling Experiment Facility for Heat Transfer Studies in Microgravity

    Delombard, Richard; McQuillen, John; Chao, David

    2008-01-01

    Pool boiling in microgravity is an area of both scientific and practical interest. By conducting tests in microgravity, it is possible to assess the effect of buoyancy on the overall boiling process and assess the relative magnitude of effects with regards to other "forces" and phenomena such as Marangoni forces, liquid momentum forces, and microlayer evaporation. The Boiling eXperiment Facility is now being built for the Microgravity Science Glovebox that will use normal perfluorohexane as a test fluid to extend the range of test conditions to include longer test durations and less liquid subcooling. Two experiments, the Microheater Array Boiling Experiment and the Nucleate Pool Boiling eXperiment will use the Boiling eXperiment Facility. The objectives of these studies are to determine the differences in local boiling heat transfer mechanisms in microgravity and normal gravity from nucleate boiling, through critical heat flux and into the transition boiling regime and to examine the bubble nucleation, growth, departure and coalescence processes. Custom-designed heaters will be utilized to achieve these objectives.

  15. Investigation of Enhanced Boiling Heat Transfer from Porous Surfaces

    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.

  16. The entrance effect on subcooled boiling in heated channels

    One of the major problems in the analysis of diabatic two-phase flows concerns the effect of thermodynamic nonequilibrium between the phases. In particular, this effect applies to forced-convection subcooled boiling in boiling water reactors (BWRs). An approach commonly used to evaluate the void distribution along reactor coolant channels is based on one-dimensional models of combined two-phase flow and boiling heat transfer. In the subcooled boiling region, the rate of phase change is governed mainly by the lateral transport of the vapor phase toward the subcooled liquid; thus, the related processes cannot be mechanistically modeled by one-dimensional, axially dependent models. Consequently, most existing subcooled boiling models are based on experimental correlations for parameters such as the onset of nucleate boiling (ONB) and the net vapor generation rate. This paper presents the results of analysis of subcooled boiling phenomena in the developing flow region of a boiling channel, based on a mechanistic two-dimensional, two-fluid model. The effect of turbulence has been accounted for by a k-ε model. The PHOENICS code was used to solve the governing mass, momentum, and energy conservation equations in both the nonboiling and boiling regions. The parameters calculated by the model include radially and axially dependent distributions of the local void fraction, temperatures and velocities of both phases, and the axial distribution of wall temperature

  17. Fuel assembly for a boiling water reactor

    A boiling water reactor fuel assembly is described which has vertical fuel rods and guide tubes positioned below the fuel rods and receiving control rod fingers and acting as water pipes, the guide tubes each being formed of a plurality of parts including a part secured to a grid plate positioned in the fuel assembly container, and low parts which fit into holes formed in the bottom of the fuel assembly. There is a flexible connection between the upper and lower parts of the guide tubes to allow for a certain tolerance in the procedure of manufacturing the various parts to allow insertion of the fuel rod bundle into the fuel assembly container

  18. Fuel recycling in boiling water reactors

    The present study confirms the feasibility of inserting mixed-oxid-fuel assemblies (MOX-FA) in boiling-water reactors in conjunction with reactivity-equivalent uranium-fuel assemblies. First, the established calculation methods were extended according to the specific MOX-uranium mutual interaction effects. Then, typical bundle-structures were analysed according to their neutron-physical features. The reactor-simulations show a non-critical behaviour with respect to limiting conditions and reactivity control. The variation of the isotopic composition and the plutonium content with its effects on the physical features was considered. (orig.) With 6 refs., 3 tabs., 29 figs

  19. Explosive Boiling of Superheated Cryogenic Liquids

    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

  20. Simulation of Boiling Water Reactor dynamics

    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)

  1. Fuel assembly for a boiling water reactor

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

  2. Study of advanced LWR cores for effective use of plutonium and MOX physics experiments

    Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ∼5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ∼6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. (author)

  3. Theoretical investigations of the transition from bubble boiling to film boiling at forced convection

    The model laws for the initial film boiling at forced convection are realized in vertical tubes. The local conditions in the investigated area were regarded to be most effective and sufficient for the description. The theory was confirmed by experimental data. (orig.)

  4. Safety implications of an integrated boiling curve model for water-cooled divertor channels

    The international fusion community is actively researching advanced heat transfer methods for removal of high thermal loads from next-generation divertor assemblies. Such advanced techniques may indeed optimize the operational and economical performance of future divertor designs. However, with its extensive operational database, water-cooling remains as one of the optimum choices for near-term divertor designs. Critical heat flux (CHF) is the maximum heat flux that water, at a given set of inlet conditions, can remove via fully developed nucleate boiling. Accordingly, an accurate CHF calculation is of the utmost importance for maintaining adequate safety margins in divertor operation. This paper uses the integrated boiling curve model developed at Sandia National Laboratories to examine the safety implications of calculating the CHF. In particular, this paper focuses on the influence of the finite element peaking factor (FEPF) that converts the heat flux predicted by CHF correlations into a plasma heat flux that can be measured. The analyses illustrate that the FEPF is proportional to the plasma heat flux and thus accurate calculation of the CHF requires the use of the appropriate FEPF for the given water conditions and plasma heat flux. It is shown that using a geometric peaking factor is inadequate since the true peak factor is dependent upon the plasma heat flux. The conclusion is that a finite element analysis incorporating an integrated boiling curve is required for accurate calculation of the CHF

  5. Nucleate Boiling from Smooth and Rough Surfaces--Part 2: Analysis of Surface Roughness Effects on Nucleate Boiling

    McHale, John P.; Garimella, Suresh V.

    2013-01-01

    The effect of surface roughness on nucleate boiling heat transfer is not clearly understood. This study is devised to conduct detailed heat transfer and bubble measurements during boiling on a heater surface with controlled roughness. This second of two companion papers presents an analysis of heat transfer and bubble ebullition in nucleate boiling with new measures of surface roughness: area ratio, surface mean normal angle, and maximum idealized surface curvature. An additional length scale...

  6. A dry-spot model of critical heat flux and transition boiling in pool and subcooled forced convection boiling

    A new dry-spot model for critical heat flux (CHF) is proposed. The new concept for dry area formation based on Poisson distribution of active nucleation sites and the critical active site number is introduced. The model is based on the boiling phenomena observed in nucleate boiling such as Poisson distribution of active nucleation sites and formation of dry spots on the heating surface. It is hypothesized that when the number of bubbles surrounding one bubble exceeds a critical number, the surrounding bubbles restrict the feed of liquid to the microlayer under the bubble. Then a dry spot of vapor will form on the heated surface. As the surface temperature is raised, more and more bubbles will have a population of surrounding active sites over the critical number. Consequently, the number of the spots will increase and the size of dry areas will increase due to merger of several dry spots. If this trend continues, the number of effective sites for heat transport through the wall will diminish, and CHF and transition boiling occur. The model is applicable to pool and subcooled forced convection boiling conditions, based on the common mechanism that CHF and transition boiling are caused by the accumulation and coalescences of dry spots. It is shown that CHF and heat flux in transition boiling can be determined without any empirical parameter based on information on the boiling parameters such as active site density and bubble diameter, etc., in nucleate boiling. It is also shown that the present model well represents actual phenomena on CHF and transition boiling and explains the mechanism on how parameters such as flow modes (pool or flow) and surface wettability influence CHF and transition boiling. Validation of the present model for CHF and transition boiling is achieved without any tuning parameter always present in earlier models. It is achieved by comparing the predictions of CHF and heat flux in transition boiling using measured boiling parameters in nucleate

  7. Unsteady heat transfer during subcooled film boiling

    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.

  8. Boiling water reactor life extension monitoring

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  9. Mitigation performance indicator for boiling water reactors

    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)

  10. Zero boil-off system testing

    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.

  11. Identification of dynamic basins in boiling fluxes

    A theoretical and experimental study of the dynamic behavior of a boiling channel is presented. In particular, the existence of different basins of attraction during instabilities was established. A fully analytical treatment of boiling channel dynamics were performed using a algebraic delay model. Subcritical and supercritical Hopf bifurcations could be identified and analyzed using perturbation methods. The derivation of a fully analytical criterion for Hopf bifurcation transcription was applied to determine the amplitude of the limit cycles and the maximum allowed perturbations necessary to break the system stability. A lumped parameters model which allows the representation of flow reversal is presented. The dynamic of very large amplitude oscillations, out of the Hopf bifurcation domain, was studied. The analysis revealed the existence of new dynamical basins of attraction, where the system may evolve to and return from with hysteresis. Finally, an experimental study was conducted, in a water loop at atmospheric pressure, designed to reproduce the operating conditions analyzed in the theory. Different dynamic phase previously predicted in the theory were found and their nonlinear characteristics were studied. In particular, subcritical and supercritical Hopf bifurcations and very large amplitude oscillations with flow reversal were identified. (author). 53 refs., figs

  12. Boiling induced mixed convection in cooling loops

    This article describes the SUCO program performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. In case of a core melt accident, the sump cooling concept realises a decay heat removal system that is based on passive safety features within the containment. The article gives, first, results of the experiments in the 1:20 linearly scaled SUCOS-2D test facility. The experimental results are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. Second, the real height SUCOT test facility with a volume and power scale of 1:356 that is aimed at investigating the mixed single-phase and two-phase natural circulation flow in the reactor sump, together with first measurement results, are discussed. Finally, a numerical approach to model the subcooled nucleate boiling phenomena in the test facility SUCOT is presented. Physical models describing interfacial mass, momentum and-heat transfer are developed and implemented in the commercial software package CFX4.1. The models are validated for an isothermal air-water bubbly flow experiment and a subcooled boiling experiment in vertical annular water flow. (author)

  13. Visualization of supercritical fluid pseudo-boiling under forced convection

    Supercritical carbon dioxide flow has been visualized by a white light to inspect forced convection heat transfer. A 'pseudo-boiling' phenomenon which occurred in supercritical range for carbon dioxide flow was supposed to cause heat transfer deterioration. The effect of the pseudo-boiling phenomenon to the heat transfer has been investigated by the visualized images in this study. (author)

  14. Direct Numerical Simulation and Visualization of Subcooled Pool Boiling

    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.

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

    Munthe Andersen, J. G.; Dix, G. E.; Leonard, J. E.; Sun, K. H.

    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....

  16. Relief valve capacity operating under conditions of working fluid boiling

    To develop a method for the calculation of relief valve capacity in the mode of working fluid boiling the effect of flowing part geometry and working fluid initial state parameters on the valve capacity is studied experimentally. Dependences for the calculation of the valve capacity in the zones of one-phase coolant flow, its boiling at high underheatings and intensive evaporation are obtained

  17. Study of neutron noise physical model for reactor coolant boiling

    The neutron noise method has been used to monitoring reactor coolant boiling. Wach-Kosaly model has been used to interpret the neutron noise induced by coolant boiling. The equation based on the model is got and used for calculation. The physical variable with the relation of bubble's velocity is got from the calculated result (autopower spectral density)

  18. Flow boiling experiment study of OTSG in steady state

    Generally, Integrated Pressurized Water Reactor uses OTSG. In this paper OTSG's flow boiling laws is analyzed and experimented. More elaborate calculating model of two-phase flow section which is divided into different segments is used for theoretical analysis. Flow boiling experiment is operated in two different conditions. The results of experiment proves the calculation model rational and credible. (authors)

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

    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

  20. Flow boiling heat transfer in volumetrically heated packed bed

    Highlights: • The onset of nucleate boiling in the volumetrically heated packed bed is researched. • A correlation for predicting qONB is developed. • The effects on boiling heat transfer coefficient are investigated. - Abstract: The volumetrically heated packed bed has been widely utilized in modern industry. However, due to the variability and randomness of packed bed channels, flow boiling heat transfer characteristics becomes complex, and there are no published research regarding this topic. To study flow boiling heat transfer characteristics of volumetrically heated packed beds, electromagnetic induction heating method is used to heat oxidized carbon steel balls adopted to stack the packed bed, with water as coolant in the experiment. The experimental results indicate that heat flux at onset of nucleate boiling (ONB) increases as mass flux and inlet subcooling are increased. A new correlation is developed to predict the ONB heat flux qONB in volumetrically heated packed bed, the predictions by which agree well with the experimental data, and the deviation remains less than 15%. Subcooled flow boiling heat transfer coefficient (hsub) increases with increasing mass flux, and equilibrium quality is slightly affected by heat flux. The saturated flow boiling heat transfer coefficient (hsat) increases with mass flux and equilibrium quality when equilibrium quality is lower than about 0.05, while the nucleate boiling is suppressed when the equilibrium quality exceeds a certain value

  1. Positive experience in the construction and project management of Kashiwazaki-Kariwa no. 6 and no. 7

    Kashiwazaki-Kariwa No.6 and No.7 (K-6/7), the world first Advanced Boiling Water Reactor (ABWR) units, started commercial operations on November 7, 1996 and July 2, 1997 respectively. ABWR has been developed as a standard design of the next generation BWR to meet common goals set by the Japanese electric utilities and BWR manufacturers (GE, Hitachi and Toshiba) based on our design, construction, operation and maintenance experience of nuclear power plants. The construction of K-6/7 and confirmatory tests for the verification of the first-of-a-kind (FOAK) design features of ABWR were conducted smoothly without any delay. The duration of the construction was 51.5 months. It was shorter than conventional BWRs in Japan by nearly one year. This was realized by design features of ABWR for better constructability, a principle of 'test before use' applied to the FOAKs, advanced construction technology, detailed engineering at very early stages of the project, and good construction management. The positive experience in the K-6/7 project is now being reviewed and standardized for next ABWR projects. The data and knowledge accumulated through the K-6/7 project will be utilized effectively with the aide of the latest information technology. (author)

  2. Bubble transport in subcooled flow boiling

    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

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

    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)

  4. Sensitivity Study for Wall Boiling Model in ANSYS CFX

    Because boiling heat transfer was crucial for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive studies have been made to develop a variety of methods either to evaluate the boiling heat transfer coefficient or to assess the onset of critical heat flux (CHF) at various operating conditions of heating channels. Because of the limitation in grid resolution for a CFD simulation in comparison with the microscopic length scales of the wall boiling process, empirical closure for some underlying physical process is needed. The main objective of the present study is to conduct the sensitivity study for wall boiling related models in ANSYS CFX R.14 in order to examine the effect of model components on wall boiling heat transfer in an axis-symmetric vertical heated pipe

  5. Mechanistic Multidimensional Modeling of Forced Convection Boiling Heat Transfer

    Michael Z. Podowski

    2009-01-01

    Full Text Available Due to the importance of boiling heat transfer in general, and boiling crisis in particular, for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive efforts have been made in the past to develop a variety of methods and tools to evaluate the boiling heat transfer coefficient and to assess the onset of temperature excursion and critical heat flux (CHF at various operating conditions of boiling channels. The objective of this paper is to present mathematical modeling concepts behind the development of mechanistic multidimensional models of low-quality forced convection boiling, including the mechanisms leading to temperature excursion and the onset of CHF.

  6. The key role of critical mock-up facilities for neutronic physics assessment of advanced reactors: an overview of Cea Cadarache tools

    The Experimental Physics section of CEA Cadarache operates three critical facilities devoted to neutronic studies of advanced reactors (EOLE, MINERVE and MASURCA) covering a large scope of interests. These include 100% MOX core in ABWR qualification, knowledge improvement of basic nuclear data for heavy nuclides for new options of the fuel cycle - especially the multi-recycling of plutonium - and accelerator-driven systems neutronic behaviour for transmutation studies. The paper describes these facilities, the scientific programmes associated and the progressive improvement of experimental techniques, the aim being to significantly reduce the uncertainties regarding the evaluation of the physical parameters. (authors)

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

    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 ...

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

    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)

  9. Self-propelled film-boiling liquids

    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.

  10. Hybrid modelling of a sugar boiling process

    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...

  11. Developments in predicting subcooled flow boiling CHF

    A two-phase flow model was developed to predict a critical heat flux (CHF) in the subcooled and low quality flow boiling. The CHF formula was derived from the local conservation equations of mass, energy and momentum, together with appropriate constitutive relations. The limiting transverse interchange of mass flux crossing the interface of the bubbly layer and core region is represented, in the local momentum conservation equation, by taking account of the convective shear effects due to the drag force on the wall-attached bubbles. Comparison between the predictions by the proposed model and the experimental CHF data from several sources shows good agreement over a wide range of flow conditions (2 P 20 MPa, 1 D 37.5 mm, 0.035 L 6 m, 450 G 7500 kg/m2s, exit 0.8). Also the model correctly accounts for the effects of flow variables

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

    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)

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

    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).

  14. Detonating gas in boiling water reactors

    The radiation in the core region of Boiling Water Reactors (BWRs) decomposes a small fraction of the coolant into hydrogen and oxygen, a phenomenon termed radiolysis. The radiolysis gas partitions to the steam during boiling. A 1000 MWe BWR produces around 1.5 tons of steam, containing 25 grams of radiolysis gas, per second. Practically all of the radiolysis gas is carried to the condenser and is taken care of by the condenser evacuation system and the off-gas system. The operation of these systems has been largely trouble-free. Radiolysis gas may also accumulate when stagnant steam condenses in pressurized pipes and components as a result of heat loss. Under certain circumstances a burnable mixture of hydrogen, oxygen and steam may form. Occasionally, the accumulated radiolysis gas has ignited. These incidents typically result in deformation of the components involved, but overpressure bursts have also occurred. Radiolysis gas accumulation in steam systems was largely overlooked by BWR designers (a likely technical reason for this is given in the report) and the problem had to be addressed by utilities. Even though the problem was recognized two decades ago, the counter-measures of today seem not always to be sufficient. Pipe-burst incidents in a German and a Japanese BWR recently attracted attention. Also, damage to a pilot valve in the steam relief system of a Swedish BWR forced a reactor shut-down during 2002. The recent incidents indicate that counter-measures against radiolysis gas accumulation in BWRs should be reviewed, perhaps also improved. The present report provides a short compilation of basic information related to radiolysis gas accumulation in BWRs. It is hoped that the compilation may prove useful to utilities and regulators reviewing the problem

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

    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

  16. Boiling water reactor simulator. Workshop material. 2. ed

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development, and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, Reactor Simulator Development: Workshop Material (2001). Course material for workshops using a WWER-1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21, 2nd edition, WWER-1000 Reactor Simulator: Workshop Material (2005). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator: Workshop Material (2005). This report consists of course material for workshops using a boiling water reactor (BWR) simulator

  17. Boiling-Water Reactor internals aging degradation study

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

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

    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.

  19. Two-Dimensional Numerical Simulation of Boiling Two-Phase Flow of Liquid Nitrogen

    Ishimoto, Jun; Oike, Mamoru; Kamijo, Kenjiro

    Two-dimensional characteristics of the boiling two-phase flow of liquid nitrogen in a duct flow are numerically investigated to contribute to the further development of new high-performance cryogenic engineering applications. First, the governing equations of the boiling two-phase flow of liquid nitrogen based on the unsteady drift-flux model are presented and several flow characteristics are numerically calculated taking account the effect of cryogenic flow states. Based on the numerical results, a two-dimensional structure of the boiling two-phase flow of liquid nitrogen is shown in detail, and it is found that the phase change of liquid nitrogen occurs in quite a short time interval compared with that of two-phase pressurized water at high temperature. Next, it is clarified that the distributions of pressure and the void fraction in a two-phase flow show a tendency different from those of fluids at room temperature because of the decrease in sound velocity due to large compressibility and the rapid phase change velocity in a cryogenic two-phase mixture flow. According to these numerical results, the fundamental characteristics of the cryogenic two-phase flow are predicted. The numerical results obtained will contribute to advanced cryogenic industrial applications.

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

    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.

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

    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.

  2. Signal processing techniques for sodium boiling noise detection

    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

  3. How DNS could help understanding basic mechanisms of boiling crisis

    Full text of publication follows: Boiling crisis characterizes the upper bound of the heat transfer (critical heat flux, CHF) between a wall and a boiling liquid at low wall superheat. This limit is associated with the departure from nucleate boiling and the establishment of stable film boiling. The nature of the crisis and the mechanisms of the transition are still objects of study. The large scope of interacting phenomena at different scales involved in boiling processes at high heat fluxes prevents from getting precise information experimentally. As a consequence, a large variety of mechanistic models has been proposed that could not be attested experimentally. The influence of the physical parameters on the value of the critical heat flux is not clearly established and this limits the predictive use of existing correlations. One could remark that all physical phenomena affecting CHF affect nucleate wall boiling as well. Recent experimental results show bubble growth, local dryness of the wall and burnout to be closely linked. One can reasonably retain the hypothesis that boiling crisis is related to a crisis during a single bubble growth event. Models considering a mechanism of local crisis, and scaling the drying process with either thermal or mechanical considerations are reviewed. This study leads to a possible scenario consistent with experimental observations. We present computations using direct numerical simulation which are intended to be a first validation of the proposed mechanism. Next steps consist in studying the interaction between the proposed mechanism and the whole boiling process and its potentiality for being the triggering phenomenon of the boiling crisis in different configurations. (authors)

  4. Enhancement of flow boiling of subcooled water on transverse ribbed surface

    The present work deals with enhancement of flow boiling heat transfer using transverse ribs of rectangular cross section attached on a flat heating surface. High flux cooling is envisioned in the field of the leading edge technology such as high power laser applications, advanced metallurgical processes, fusion reactors, integrated circuit chips and so forth. In such advanced cooling devices subcooled boiling of water at high velocity is expected as one of most efficient and convenient means for heat removal exceeding 107W/m2. In the present experiment, a sheet of stainless-steel (10mm wide, 0.2mm thick and 80mm in heated length) was flush mounted on one wall of a vertical rectangular channel (a cross-section 20mm x 30 mm) and used as a heating surface by passing a direct current. Transverse-ribs made of plastic plate (0.5mm thick and 30mm in lateral length) were attached at an equal longitudinal spacing on the heating surface. Longitudinal spacing of ribs was varied from 2.5, 5, 10, 20mm to infinity (a flat surface without ribs), and the rib height was 2.5mm and 5.0mm. Experiments were conducted with water at a pressure of 0.12MPa in the range of mass velocity from 500kg/m2s to 2,000kg/m2s (water velocity from 0.5m/s to 2m/s) and subcooling from 20K to 50K. It was found that the ribbed structure strongly affects heat transfer both in the non-boiling and partial nucleate boiling regimes. Enhancement rate of heat transfer coefficient varied from 5% to 50% in excess compared with that for the flat heating plate. According to visual observation of boiling bubbles the circulating flow occurred in a space between each consecutive ribs and seemed to enhance heat transfer. For narrow rib spacing a large coalesced bubble filled up each rib space and impeded the exchange of vapor with liquid, leading to heat transfer deterioration. In the present experimental range of the ribs, a 50% increase of heat transfer coefficient was attained on the ribbed surface with 5mm height

  5. Strategy of plant concept development in Hitachi

    Hitachi contributes to the society in various fields of nuclear power such as the Light Water Reactor field, the Fast Breeder Reactor, the fuel cycle, and the medical treatment. Since the beginning of a first commercial operation of a BWR in Japan, Hitachi has constructed 20 units of BWR. Hitachi continues its efforts in achieving high reliability and large-scale output, and in 1996, it completed Advanced Boiling Water Reactor (ABWR) in cooperation with various BWR utilities, General Electric Company, and Toshiba Corporation. Hitachi has enhanced the ABWR technology further based on the above enough experience. The latest technologies were reflected in Hamaoka unit 5 and Shika unit 2 as the latest ABWR plants. The further upgrade technologies would been reflected in Shimane-3, Ohma-1 and Higashidori-1 as ABWR plants under planning. Hitachi obtains the chance to construct the nuclear power plant continuously. For the next generation, Hitachi is working on developing nuclear power plants that take diversified needs and global characteristics into account. As one of the approaches, the output series formation of BWR is extended as the following. ABWR-II and ESBWR (Economic and Simplified BWR) as a large-scale centralized power supply emphasizing cost efficiency, Medium size ABWR and natural circulation type BWR as medium and small-scale distributed power supply that features flexibility to various market needs, such as minimized capital risks, timely return on a capital investment, etc. As another approach, Hitachi tries to extend the Light Water Reactor technology. RBWR (Resource-Renewable BWR ) that achieves a high conversion ratio over 1.0 based on the BWR technology will make the fuel cycle flexible. Hitachi will continue the challenge for the next ABWR and the future with the enough experience of BWR construction. (author)

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

    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.

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

    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...

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

    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)

  9. Boiling Heat Transfer on Porous Surfaces with Vapor Channels

    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.

  10. Study on instability of natural circulation induced by subcooled boiling

    The best estimate system analysis code RELAP5 was used to analyze the natural circulation systems. The instability boundaries of one natural circulation system were obtained under different conditions. According to present results, most of the boundary points were found in the low subcooled boiling zone. The natural circulation systems can tolerate high subcooled boiling, and the disturbance of bubbles departing from the wall and condensing in the subcooled boiling region may be the inherent source to induce the instability, then the flow oscillations can become self-sustained and evolve because of the phase differences among system driving force, resistance and flow rate. (authors)

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

    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

  12. Critical heat flux in subcooled flow boiling

    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

  13. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

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

    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…

  15. 40 CFR 180.1056 - Boiled linseed oil; exemption from requirement of tolerance.

    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....

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

    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.

  17. Prediction of transition boiling heat transfer by artificial neural network

    Based on the capability of nonlinear mapping of artificial neural network, a neural network is presented to predict the transition boiling heat transfer in vertical annulus and vertical tube. The predicting results show good accordance with the experimental results

  18. Numerical model of post-DNB film boiling heat transfer

    It is proposed in this paper a physical model for the film boiling heat transfer. The corresponding mathematical descriptions are given in details and the heat transfer characteristic of post-DNB film boiling is analyzed. The numerical model of post-DNB film boiling heat transfer is obtained as the empirical value of the coefficient is determined by the experimental data. The numerical model is compared with the experimental data of different parameters and other numerical models, and the statistical deviations are calculated. The calculating results of the numerical model in this paper show good agreement with the experimental data, and the numerical model in this paper has comprehensive applicability compared with other numerical models. The effects of thermal-hydraulic parameters on the post-DNB film boiling heat transfer have been numerically researched using the numerical model in this paper. The calculating results are as same as the experimental results. (authors)

  19. Heat transfer phenomena related to the boiling crisis

    This report contains a state-of-the-art review of critical heat flux (CHF) and post-CHF heat transfer. Part I reviews the mechanisms controlling the boiling crisis. The observed parametric trends of the CHF in a heat flux controlled system are discussed in detail, paying special attention to parameters pertaining to nuclear fuel. The various methods of predicting the critical power are described. Part II reviews the published information on transition boiling and film boiling heat transfer under forced convective conditions. Transition boiling data were found to be available only within limited ranges of conditions. The data did not permit the derivation of a correlation; however, the parametric trends were isolated from these data. (author)

  20. Transition from boiling to two-phase forced convection

    The paper presents a method for the prediction of the boundary points of the transition region between fully developed boiling and two-phase forced convection. It is shown that the concept for the determination of the onset of fully developed boiling can also be applied for the calculation of the point where the heat transfer is effected again by the forced convection. Similarly, the criterion for the onset of nucleate boiling can be used for the definition of the point where boiling is completely suppressed and pure two-phase forced convection starts. To calculate the heat transfer coefficient for the transition region, an equation is proposed that applies the boundary points and a relaxation function ensuring the smooth transition of the heat transfer coefficient at the boundaries

  1. Zero Boil Off System for Cryogen Storage Project

    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...

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

    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.)

  3. Comparative analysis of heat transfer correlations for forced convection boiling

    A critical survey was conducted of the most relevant correlations of boiling heat transfer in forced convection flow. Most of the investigations carried out on partial nucleate boiling and fully developed nucleate boiling have led to the formulation of correlations that are not able to cover a wide range of operating conditions, due to the empirical approach of the problem. A comparative analysis is therefore required in order to delineate the relative accuracy of the proposed correlations, on the basis of the experimental data presently available. The survey performed allows the evaluation of the accuracy of the different calculating procedure; the results obtained, moreover, indicate the most reliable heat transfer correlations for the different operating conditions investigated. This survey was developed for five pressure range (up to 180bar) and for both saturation and subcooled boiling condition

  4. Theory of hydraulic stability of boiling channels

    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.)

  5. Burnout heat flux in natural flow boiling

    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

  6. Characteristics of phenomenon and sound in microbubble emission boiling

    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)

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

    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...

  8. Hysteresis of boiling for different tunnel-pore surfaces

    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...

  9. Flow Boiling in straight heated tube under microgravity conditions

    Narcy, Marine; COLIN, Catherine

    2013-01-01

    Boiling two-phase flow can transfer large heat fluxes with small driving temperature differences, which is of great interest for the design of high-performance thermal management systems applied to space platforms and on-board electronics cooling in particular. However, such systems are designed using ground-based empirical correlations, which may not be reliable under microgravity conditions. Therefore, several two-phase flow (gas-liquid flow and boiling flow) experiments have been conducted...

  10. Modeling acid-gas generation from boiling chloride brines

    Sonnenthal Eric; Spycher Nicolas; Zhang Guoxiang; Steefel Carl

    2009-01-01

    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...

  11. Adiabatic boiling of two-phase coolant in upward flow

    A mathematical model of the process of adiabatic boiling (self-condensation) of a two-phase coolant in upward (downward) flow is developed. The model takes account of changes in phase properties with static pressure decrease. The process is investigated numerically. Approximate analytical formulas for design calculations are obtained. It is shown that effects of adiabatic boiling (self-condensation) should be taken into account when calculating two-phase coolant flow in stretched vertical channels

  12. Development of an experimental apparatus for nucleate boiling analysis

    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)

  13. Thermodynamic crisis in boiling flow. Observation of the flicker noise

    The results of the experimental studies on both the characteristics of the boiling liquid flow (discharge, jet reactive force), emanating through a short channel, and the local pulsations in the flow are presented. The identified effects - the flow critical mode, sharp decrease in the value of the reactive force, pulsations with the 1/f spectrum (the flicker noise) are discussed with attraction of the notion on the boiling thermodynamic crisis

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

    A. V. Оvsiannik

    2014-01-01

    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 ...

  15. Visualization of pool boiling on downward-facing convex surfaces

    Visualizations and quenching experiments were performed to investigate effect of material properties on pool boiling from downward-facing, convex stainless steel and copper surfaces in saturated water. Video images showed that more than one boiling regimes can co-exist on the surface. Maximum heat flux (MHF) occurred first at lowermost position, then propagated radially outward to higher inclination positions and its local value decreased with increased inclination. However, the wall superheats corresponding to MHF were independent of the local surface inclinations. MHF propagated ∼10 times slower on stainless-steel than on copper and was ∼12% and 40% lower on stainless-steel than on copper at θ = 0 degree and θ 7.91 degree, respectively. Results confirmed that transition boiling consisted of two distinct regions: high wall superheat, in which heat flux increased relatively slowly, and low wall superheat, in which heat flux increased precipitously with time. Nuclear boiling regime also consisted of two distinct regions: high heat flux nucleate boiling, in which heat flux decreased with increased inclination, and low heat flux nucleate boiling, in which heat flux increased with increased inclination

  16. Contamination of Foods Boiled in Containers of Different Materials

    The data presented in this communication deals with the contamination of milk and tomatoes juice, that are a daily consumed by a broad spectrum of Egyptian population when boiled in different types of utensils namely, glass, tin old and new aluminum, pottery, tefal, and stainless steel with heavy metals. The elements determined were sodium, calcium, potassium, iron, manganese, zinc, strontium, cobalt, cadmium, mercury, tin, rubidium, ytterbium and antimony. The technique used for the simultaneous determination of these elements was the instrumental neutron activation analysis. In the light of the obtained results, it was suggested that glass utensils are preferred for boiling milk on the others types of utensils. Also it was found that there is no danger when milk is boiled in other types of utensils since the concentration of both essential and hazardous metals were in the tolerable range. In case of boiling tomatoes juice in the aforementioned types of utensils, it was found that there was no distinct difference between the elemental content in the samples boiled in those types of utensils, thus it can be deduced that all these types of utensils can be used safety for boiling or cooking tomatoes juice

  17. Coupling of wall boiling with discrete population balance model

    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)

  18. Film boiling of R-11 on liquid metal surfaces

    An interesting problem is the effect of an immiscible liquid heating surface on the process of film boiling. Such surfaces raise questions concerning interface stability to disturbances, effects of gas bubbling, and vapor explosions in layered systems. The specific motivation for this study was to investigate film boiling from a liquid surface with application to cooling of molten reactor core debris by an overlying pool of reactor coolant. To investigate this phenomenon, and apparatus consisting of a nominal six-inch diameter steel vessel to hold the liquid metal and boiling fluid was constructed; coolant reservoirs, heaters, controllers, and allied instrumentation were attached. A transient energy balance was performed on the liquid metal pool by a submerged assembly of microthermocouples in the liquid metal and an array of thermocouples on the wall of the test vessel. The thermocouple data were used to determine the boiling heat flux as well as the boiling superheat. On an average basis, the deviation between the prediction of the Berenson model and the experimental data was less than one percent when Berenson was corrected for thermal radiation effects. Evidence from visualization tests of R-11 in film boiling over molten metal pools to superheats in excess of 600 K supports this conclusion. 13 refs

  19. Experiments of Pool Boiling Performance (Boiling Heat Transfer and Critical Heat Flux) on Designed Micro-Structures

    In general, the evaluation of the boiling performance mainly focuses on two physical parameters: boiling heat transfer (BHT) and critical heat flux (CHF). In the nuclear power plants, both BHT and CHF contribute the nuclear system efficiency and safety, respectively. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on Pin-fin effect analysis. In terms of CHF, critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on the roughness ratio. The extended heat transfer area contributes the boiling heat transfer increase on the structured surface, and its quantitative analysis has been performed. In terms of CHF, the critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. We suggested a capillary limit to CHF delay for modeling capillary induced liquid inflow through microstructured surfaces. The critical size of the capillary limit on the prepared structured surface, determined by a model, could be reasonable explanation points for the experimental results (optimal size for CHF delay). The present experimental results also showed clearly the critical size (10 - 20 μm) for CHF delay, predicted by capillary limit analysis. This study provides fundamental insight into BHT and CHF enhancement of structured surfaces, and an optimal design guide for the required CHF and boiling heat-transfer performance. Finally, this study can contribute the basic understanding of the boiling on designed microstructure surface, and it also suggest the optimal micro scaled structured surface of boiling

  20. Development of sodium boiling model, 'SOBOIL'

    The objective of this research is to develop an algorithm for the sodium boiling modeling, essential to the KALIMER HCDA Analysis. The basic theory is based on the 'Muti-slug Ejection Model' in SAS2A. It allows a finite number of bubbles in a channel at any time, and a formed bubble fills the whole cross section of the coolant channel except for liquid film left on the cladding surface. The essence of the model is to estimate the bubble pressure and temperature by balancing the bubble state change with the energy transfer into the bubble at both the wall and interfaces. It assumes that the bubble is saturated with a uniform pressure and temperature, and a bubble is generated when the coolant temperature exceeds a specified superheat. The period of the algorithm development has been divided into two phases. The algorithm development and application during the first phase is limited to the active fuel region where relatively simple physical phenomena are anticipated under ULOHS accident conditions in KALIMER, because it is favorable to the verification of the basic algorithm. The main revision is made during the 2nd phase to take into account mass transfer between the liquid film and bubble, bubble collapse, and coalescence as a liquid slug diminishes. In conclusion, the model represents the anticipated physical phenomena reasonably. The bubble has grown rapidly in consistence with the previous model predictions. Instability, however, is found in the wall heat transfer because the bubble size exceeds a certain value, over which the homogeneous model is no longer valid. Therefore, the model is expected to be improved by taking account of the pressure drop due to vapor flow inside such a large bubble to extend its applicability

  1. Study of film boiling collapse behavior during vapor explosion

    Possible large scale vapor explosions are safety concern in nuclear power plants during severe accident. In order to identify the occurrence of the vapor explosion and to estimate the magnitude of the induced pressure pulse, it is necessary to investigate the triggering condition for the vapor explosion. As a first step of this study, scooping analysis was conducted with a simulation code based on thermal detonation model. It was found that the pressure at the collapse of film boiling much affects the trigger condition of vapor explosion. Based on this analytical results, basic experiments were conducted to clarify the collapse conditions of film boiling on a high temperature solid ball surface. Film boiling condition was established by flooding water onto a high temperature stainless steel ball heated by a high frequency induction heater. After the film boiling was established, the pressure pulse generated by a shock tube was applied to collapse the steam film on the ball surface. As the experimental boundary conditions, materials and size of the balls, magnitude of pressure pulse and initial temperature of the carbon and stainless steel balls were varied. The transients of pressure and surface temperature were measured. It was found that the surface temperature on the balls sharply decreased when the pressure wave passed through the film on balls. Based on the surface temperature behavior, the film boiling collapse pattern was found to be categorized into several types. Especially, the pattern for stainless steel ball was categorized into three types; no collapse, collapse and reestablishment after collapse. It was thus clarified that the film boiling collapse behavior was identified by initial conditions and that the pressure required to collapse film boiling strongly depended on the initial surface temperature. The present results will provide a useful information for the analysis of vapor explosions based on the thermal detonation model. (J.P.N.)

  2. Inspection of Pool Boiling with Superhydrophilic and Superhydrophobic Coating

    Son, Gyumin; Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    In conventional nuclear power plants, increasing critical heat flux (CHF) margin by converting existing parts is economically meaningful since it means overall energy production increase without building additional power plants. There were researches to enhance margin from the very beginning of the commercialization of nuclear power plants and many efforts have led to current model of plants, optimized for both safety and production efficiency. Examples are mixing vane which is actually applied to plants nowadays, using nanofluids to enhance heat transfer coefficient (HTC), trying porous surfaces and so on. Takata et al. studied effects of surface wettability by using hydrophobic coating and observed enhanced nucleate boiling at coated surface regions. Betz et al. experimented superhydrophilic (SHPi), superhydrophobic (SHPo), and superbiphilic surfaces. Results indicate heat transfer coefficient enhancement due to increase of nucleation sites by hydrophobic regions and constrained diameter of growing bubbles by hydrophilic regions. Although it would be rough to apply their concept to real reactor coolant surface wall, understanding the possibility of enhanced boiling is meaningful. In this paper, SHPi and SHPo coatings were applied to wire at traditional pool boiling experiment by Nukiyama. By observing altered CHF margin and nucleate boiling, the effects of each coating and their tendencies are discussed. SHPi, SHPo and bare wire's pool boiling was investigated and their boiling graphs were discussed. SHPi shows enhancement in CHF while SHPo's case is more complicated since there were variables like partial CHF or micro scale bubbles. Additional experiment could be comparing HTC, checking whether hydrophobic wire's nucleate boiling enhancement can exceed the decreased CHF margin. More sophisticated method to remove unwanted bubbles should be considered such as using degassed water.

  3. Study of film boiling collapse behavior during vapor explosion

    Yagi, Masahiro; Yamano, Norihiro; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Abe, Yutaka; Adachi, Hiromichi; Kobayashi, Tomoyoshi

    1996-06-01

    Possible large scale vapor explosions are safety concern in nuclear power plants during severe accident. In order to identify the occurrence of the vapor explosion and to estimate the magnitude of the induced pressure pulse, it is necessary to investigate the triggering condition for the vapor explosion. As a first step of this study, scooping analysis was conducted with a simulation code based on thermal detonation model. It was found that the pressure at the collapse of film boiling much affects the trigger condition of vapor explosion. Based on this analytical results, basic experiments were conducted to clarify the collapse conditions of film boiling on a high temperature solid ball surface. Film boiling condition was established by flooding water onto a high temperature stainless steel ball heated by a high frequency induction heater. After the film boiling was established, the pressure pulse generated by a shock tube was applied to collapse the steam film on the ball surface. As the experimental boundary conditions, materials and size of the balls, magnitude of pressure pulse and initial temperature of the carbon and stainless steel balls were varied. The transients of pressure and surface temperature were measured. It was found that the surface temperature on the balls sharply decreased when the pressure wave passed through the film on balls. Based on the surface temperature behavior, the film boiling collapse pattern was found to be categorized into several types. Especially, the pattern for stainless steel ball was categorized into three types; no collapse, collapse and reestablishment after collapse. It was thus clarified that the film boiling collapse behavior was identified by initial conditions and that the pressure required to collapse film boiling strongly depended on the initial surface temperature. The present results will provide a useful information for the analysis of vapor explosions based on the thermal detonation model. (J.P.N.)

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

    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

  5. Boiling crisis and non-equilibrium drying transition

    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...

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

    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)

  7. Film boiling characteristics of potassium droplets on heated plate

    For providing background information on the possible vapor explosion in the event of a core disruptive accident of LMFBRs, an experiment was conducted on the film boiling characteristics of liquid metal potassium in association with the Leidenfrost phenomenon. In a steel container filled with Ar gas, K droplets were put on a joule-heated plate of 316-SS or Ta. The behaviors of droplet were observed by a camera and a color VTR through viewports. The experimental conditions were the Ar pressure 1 bar, the initial K temperature 350 -- 7600C, and the plate temperature 900 -- 1,2500C for 316-SS and 1,100 -- 1,6000C for Ta. Stable film boiling known as Leidenfrost phenomenon was observed for a high temperature condition of the plate, whereas an instantaneous break-up of droplet with extensive vaporization occurred for a low temperature. The heat transfer characteristics of film and transition boiling regions were obtaind by estimating the heat flux from the volumetric reducing rate of droplet due to vaporization. The results in the film boiling region showed an appreciably good agreement with the prediction based on Bromley's expression combined with the theory of Baumeister et al. The minimum film boiling temperature and heat flux were found to be about 1,3000C and 15 W/cm2, respectively, for a droplet size of 0.15 cm3. (author)

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

    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.

  9. Enhanced boiling heat transfer in horizontal test bundles

    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.

  10. Study on saturated flow boiling heat transfer under vibration conditions

    The ability to predict void formation, void fraction and critical heat flux -CHF- in flow boiling under oscillatory flow and vibration conditions is important to the safety technology of nuclear reactor during earthquake. In the present study, the onset of nucleate boiling -ONB-, the point of net vapor generation -NVG- and CHF on saturated flow boiling under vibration conditions were investigated experimentally. Steady state experiments were conducted using a copper thin-film and subcooled water at 0.1 MPa. The liquid velocity was 0.27, 1.38, 3.20 and 4.07 m/s, respectively; the liquid subcooling was 0 K. A heater was made of a printed circuit board. A test section was a rectangular flow channel of 10 mm width and 10 mm height. The test heater was heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The test section was arranged for horizontal position facing upward and for vertical position, respectively. For the vibration condition, the test section was set on a vibration table. The ONB was decided as an occurrence of the first boiling bubble. The critical heat flux was determined as that immediately before the heating surface physically burned-out. The CHF on saturated flow boiling under vibration conditions were investigated experimentally. (author)

  11. Wall function approach for boiling two-phase flows

    One of the important goals of the NURESIM project is to assess and improve the simulation capability of the three-dimensional two-fluid codes for prediction of local boiling flow processes. The boiling flow is strongly affected by local mechanisms in the turbulent boundary layer near the heated wall. Wall-to-fluid transfer models for boiling flow with the emphasis on near-wall treatment are being addressed in the paper. Since the computational grid of the 3D two-fluid models is too coarse to resolve the variable gradients in the near-wall region, the use of wall functions is a common approach to model the liquid velocity and temperature profile adjacent to the heated wall. The wall function model for momentum, based on the surface roughness analogy has been discussed and implemented in the NEPTUNECFD code. The model has been validated on several upward boiling flow experiments, differing in the geometry, working fluid and operating conditions. The simulations with the new wall function model show an improved prediction of flow parameters over the boiling boundary layer. Furthermore, a wall function model for the energy equation, based on enhanced two-phase wall friction has been derived and validated.

  12. Heater size effect on subcooled pool boiling of FC-72

    Extensive research has been conducted on pool boiling using heaters larger than the capillary length. For large heaters and/or high gravity conditions, boiling is dominated by buoyancy, and the heat transfer is heater size independent. Much less is known about boiling on small heaters and at low gravity levels. The ratio of heater size Lh to capillary length Lc is an important parameter in the determination of heater size dependence on heat transfer. As the ratio Lh/Lc decreases due to a decrease in either heater size or gravity, surface tension forces become dominant. It is proposed that transition from buoyancy to surface tension dominated boiling occurs when the heater size and bubble departure diameter are of the same order. Previous work in variable gravity with flat surfaces has shown that the heat transfer was heater size independent only when the ratio Lh/Lc was considerably larger than 1. An array of 96 platinum resistance heater elements in a 10 x 10 configuration with individual elements 0.7 x 0.7 mm2 in size was used to vary heater size and measure the heat transfer. The threshold value of Lh/Lc above which pool boiling is heater size independent was found to be about 2.8. (author)

  13. Simulation of DEBORA experiments for subcooled boiling flow by the EAGLE code with one group interfacial area transport equation

    In this study, the capability of the EAGLE code, which is an in house CFD code for two phase flow, to simulate local boiling flow processes over a wide range of operating conditions, including those close to CHF, has been assessed. The implementation of interfacial area transport (IAT) equation with advanced models such as bubble coalescence, breakup and nucleation site density is indispensable for predicting accurately the bubble size distribution. The DEBORA experiments, performed at CEA using dichlorodifluoro-methane (R12) as the working fluid, were selected to analyze a subcooled boiling flow. The aim of this work is to investigate the applicability of the state of the art physical models for interfacial area concentration (IAC)

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

    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.

  15. Changes of enthalpy slope in subcooled flow boiling

    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.)

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

    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.

  17. The concept and application of miniaturization boiling in cooling system

    The purpose of this research is to study and examine the phenomena of miniaturization-boiling, which intensely scatters with a large number of minute liquid particles from a water droplet surface to the atmosphere, when the droplet collided with a heating surface. As the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr=17%,Ni=8%), sapphire (Al3O2), brass, copper and carbon plane. The material was heated in order to study the miniaturization-boiling and droplet bounding phenomena at a very high temperature (160 degree C- 420 degree C). The phenomenon was photographed by a high-speed camera (10,000 fps) from the horizontal direction. The nuclear fusion reactor needs a very severe cooling, heat removal cooling method by special boiling is lead to this research. (Author)

  18. The concept and application of miniaturisation boiling in cooling system

    The purpose of this research is to study and examine the phenomena of miniaturisation-boiling, which intensely scattered into a large number of minute liquid particles from a water droplet surface into the atmosphere, when the droplet collided with a heating surface. For the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr= 17 %, Ni= 8 %), sapphire (Al2O3), brass, copper and carbon plane. The material was heated in order to study the miniaturisation-boiling and droplet bounding phenomena at a very high temperature (160 degree Celsius- 420 degree Celsius). The phenomenon was photographed by a high-speed camera (10000 fps) from the horizontal direction. The nuclear fusion reactor which needs a severe cooling, and heat removal cooling method through special boiling leads to this research. (author)

  19. On the dynamics of bubbles in boiling water

    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.

  20. Boiling heat transfer in porous media composed of particles

    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)

  1. Numerical model of post-DNB transition boiling heat transfer

    In this paper a physical model for the transition boiling heat transfer is proposed. The corresponding mathematical descriptions are given in detail and the heat transfer characteristics of post-DNB transition boiling is analyzed. The numerical model of post-DNB transition boiling heat transfer is obtained as the empirical value of the coefficient is determined by the experimental data. The numerical model is compared with the experimental data of different parameters and other numerical models, and the statistical deviations are calculated. The calculating results of the numerical model in this paper show good agreement with the experimental data and the numerical model in this paper is with good applicability compared with other numerical models. (authors)

  2. Evaluation of N fuels boiling subchannel flow and DNB behavior

    An updated evaluation of the boiling and burnout performance of the N Reactor fuels is presented. Several new procedures and improvements for fuel analyses are described. These include the following items incorporated into the BOTHER computer program: subroutine FLOBAL which balances the flows among the boiling and nonboiling channels to equilibrate pressure drop; a new enthalpy imbalance factor for the MKIV fuel of 19.4% (95/95 basis); an updated method for calculating the two-phase pressure drop multiplier for each channel type, which includes both momentum and friction terms; and changes in the BOTHER code. The net effect of these changes is to predict DNBR limits before connecter boiling and at slightly higher flows and lower tube powers than predicted previously

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

    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.

  4. Modeling of subcooled boiling in the vertical flow

    A two-dimensional model of subcooled boiling in a vertical channel was developed. Its basic idea is that the vapor phase generation has a similar effect on the flow field as a hypothetical liquid phase generation. The bubble volume, generated due to evaporation process, was filled with liquid and included as a source term in the continuity equation for the liquid phase. Thus, the single-phase from of transport equations was preserved and bubbles were retained in the boundary layer near the heated surface. Time development of subcooled boiling was simulated and effects of governing physical mechanisms (evaporation, condensation, vapor-phase convection, vapor-phase diffusion) on the flow field and pressure drop were analyzed. The Results of the proposed two-dimensional model were compared with experimental data and RELAP5/MOD3.2 calculations. The presented model represents a contribution to the two-dimensional simulation of the subcooled boiling phenomenon.(author)

  5. Pin cooling and dryout in steady local boiling

    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.)

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

    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.)

  7. Pool boiling from downward-facing curved surfaces: Effects of radius of curvature and edge angle

    Transient pool boiling from downward-facing curved surfaces in water is of interest for assessing the coolability of the lower head of an advanced light water reactor (ALWR) pressure vessel following a core meltdown accident. Here, quenching experiments were performed to investigate the effects of radius of curvature and edge angle on pool boiling from downwards-facing surfaces in saturated power. The experiments employed two, 20-mm-thick copper test sections that had the same diameter (75 mm) but different surface radii (148 and 218.5 mm) and vapor release (or edge) angles (14.68 and 9.88 deg). The effect of surface area on pool boiling was determined by comparing the present results with the results for a copper section that was of the same thickness but had a surface radius of 148 mm and was less than one-half the surface area. The maximum heat flux (qMHF) was highest at the lowermost position and decreased with increased local inclination on the surface. Both local and surface average qMHF were representative of quasi-steady-state critical heat flux. The high edge angle reduced vapor accumulation, which enhanced surface coolability and shortened its quenching time. For an edge angle of 9.88 deg, increasing the surface area (or surface radius) insignificantly affected the local qMHF near the edge of the copper section but lowered it everywhere else by ∼10%. For the same surface area, the larger edge angle (or smaller surface radius) increased qMHF by as much as 40%

  8. Heat transfer correlation for saturated flow boiling of water

    The saturated flow boiling heat transfer of water (H2O, R718) is encountered in many applications such as compact heat exchangers and electronic cooling, for which an accurate correlation of evaporative heat transfer coefficients is necessary. A number of correlations for two-phase flow boiling heat transfer coefficients were proposed. However, their prediction accuracies for H2O are not satisfactory. This work compiles an H2O database of 1055 experimental data points from micro/mini-channels from nine independent studies, evaluates 41 existing correlations to provide a clue for developing a better correlation of saturated flow boiling heat transfer coefficients for H2O, and then proposes a new one. The new correlation incorporates a newly proposed dimensionless number and makes great progress in prediction accuracy. It has a mean absolute deviation of 10.1%, predicting 81.9% of the entire database within ±15% and 91.2% within ±20%, far better than the best existing one. Besides, it also works well for several other working fluids, such as R22, R134a, R410A and NH3 (ammonia, R717), being the best for R22, R410A and NH3 so far. - Highlights: • Compiles a database of 1055 data points of H2O flow boiling heat transfer. • Evaluates 41 correlations of flow boiling heat transfer coefficient. • Generalize approach for developing experiment-based correlation. • Propose a correlation of H2O flow boiling heat transfer in small channels. • The new correlation has a mean absolute deviation of 10.1% for the database

  9. Saturated Pool Boiling in Vertical Annulus with Reduced Outflow Area

    The mechanisms of pool boiling heat transfer have been studied extensively to design efficient heat transfer devices or to assure the integrity of safety related systems. However, knowledge on pool boiling heat transfer in a confined space is still quite limited. The confined nucleate boiling is an effective technique to enhance heat transfer. Improved heat transfer might be attributed to an increase in the heat transfer coefficient due to vaporization from the thin liquid film on the heating surface or increased bubble activity. According to Cornwell and Houston, the bubbles sliding on the heated surface agitate environmental liquid. In a confined space a kind of pulsating flow due to the bubbles is created and, as a result very active liquid agitation is generated. The increase in the intensity of liquid agitation results in heat transfer enhancement. Sometimes a deterioration of heat transfer appears at high heat fluxes for confined boiling. The cause of the deterioration is suggested as active bubble coalescence. Recently, Kang published inflow effects on pool boiling heat transfer in a vertical annulus with closed bottoms. Kang regulated the gap size at the upper regions of the annulus and identified that effects of the reduced gaps on heat transfer become evident as the heat flux increases. This kind of geometry is found in an in-pile test section. Since more detailed analysis is necessary, effects of the outflow area on nucleate pool boiling heat transfer are investigated in this study. Up to the author's knowledge, no previous results concerning to this effect have been published yet

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

    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)

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

    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.

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

    Yazdani, Miad; Radcliff, Thomas; Soteriou, Marios; Alahyari, Abbas A.

    2016-01-01

    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.

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

    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

  14. Immersion cooling nucleate boiling of high power computer chips

    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.

  15. Use of Synchronized, Infrared Thermometry and High-Speed Video for Generation of Space- and Time- Resolved High-Quality Data on Boiling Heat Transfer

    Nucleate boiling is an effective mode of heat transfer; one of the most studied physical phenomena in science and engineering, and a key thermal limit in nuclear systems. However, for decades, modeling of nucleate boiling heat transfer has been relying on speculative hypotheses and a good dose of empiricism. For example, the widely popular Rosenhow's correlation for nucleate boiling is based on the assumption that single-phase convection and nucleate boiling are analogous physical processes, and can be both correlated in terms of the Reynolds and Prandtl number of the liquid phase; for nucleate boiling the characteristic velocity and length are assumed to be the downward liquid velocity and most unstable Taylor wavelength, respectively; then, an empirical constant, Csf, is determined to fit the experimental data for any fluid/surface combination. As researchers are now finally moving away from the rough empiricism of the past, and start to develop more mechanistic models of nucleate boiling heat transfer, the need for high-quality high-resolution data on the bubble nucleation and growth cycle is becoming increasingly big. Specifically, nucleation site density, bubble departure diameter and frequency data are necessary input for the source terms in interfacial area transport models and CFD 'multi-fluid' models as well as semi-empirical models for boiling heat transfer, such as the RPI's heat flux partitioning model and Kolev's bubble interaction model. Furthermore, time-resolved temperature distribution data for the boiling surface and direct visualization of the bubble cycle are needed for validation of 'first principle' models of bubble nucleation and growth, based on interface tracking methods, in which the geometry of the vapor/liquid interface is not assumed, but rather calculated from a marker function advected according to the Navier-Stokes equations. However, gathering the detailed data needed for validation of advanced simulation models is not

  16. A stability identification system for boiling water nuclear reactors

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  17. Hysteresis of boiling for different tunnel-pore surfaces

    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.

  18. On Boiling of Crude Oil under Elevated Pressure

    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.

  19. On Boiling of Crude Oil under Elevated Pressure

    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.

  20. Dimensional analysis of boiling heat transfer burnout conditions

    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

  1. Prediction of void fraction in subcooled flow boiling

    The information on heat transfer and especially on the void fraction in the reactor core under subcooled conditions is very important for the water-cooled nuclear reactors, because of its influence upon the reactivity of the systems. This paper gives a short overview of subcooled boiling phenomenon and indicates the simplifications made by the RELAP5 model of subcooled boiling. RELAP5/MOD3.2 calculations were compared with simple one-dimensional models and with high-pressure Bartolomey experiments.(author)

  2. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    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)...

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

    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.

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

    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...

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

    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...

  6. Evaluation of Correlations of Flow Boiling Heat Transfer of R22 in Horizontal Channels

    Zhanru Zhou; Xiande Fang; Dingkun Li

    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 corr...

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

    Mr.P. Atcha Rao; Mr.V.V.Ramakrishna

    2015-01-01

    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 surf...

  8. 46 CFR 154.705 - Cargo boil-off as fuel: General.

    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...

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

    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. Research on radiation detectors, boiling transients, and organic lubricants

    1974-01-01

    The accomplishments of a space projects research facility are presented. The subjects discussed are: (1) a study of radiation resistant semiconductor devices, (2) synthesis of high temperature organic lubricants, (3) departure from phase equilibrium during boiling transients, (4) effects of neutron irradiation on defect state in tungsten, and (5) determination of photon response function of NE-213 liquid scintillation detectors.

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

    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.

  12. Film boiling on downward quenching hemisphere of varying sizes

    Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Laminar Transition Apparatus). Two test sections are made of copper to maintain low Biot numbers. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of all the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the test sections. The measured data are compared with the numerical predictions from laminar film boiling analysis. The measured heat transfer coefficients are found to be greater than those predicted by the conventional laminar flow theory on account of the interfacial wavy motion incurred by the Helmholtz instability. Incorporation of the wavy motion model considerably improves the agreement between the experimental and numerical results in terms of heat transfer coefficient. In addition, the interfacial wavy motion and the quenching process are visualized through a digital camera. (authors)

  13. Investigations on coolant boiling in research reactors. 2

    Subcooled boiling has been investigated systematically at the Rossendorf Research Reactor in the range between boiling onset and boiling crisis. This is of particular interest because in the core the direction of the coolant flow is opposite to the bubble buoyance of the bubbles - in contrast to power reactors. For this reason an experimental fuel assembly equipped with a throttle valve for coolant flow reduction and different detectors was built up and installed in the reactor core. Measurements of thermohydraulic parameters and noise signals from temperature, neutron flux and acoustic sources were subject of the investigations. Besides other results fluctuations of the void fraction induced by a standing wave of the two-phase flow in the coolant channel and the 24-Hz pressure fluctuations of the circulation pumps have been observed. It has been shown that the frequency of the standing wave is determined by the size of the boiling volume in the coolant channel and that this frequency therefore depends on the outlet temperature of the coolant. (author)

  14. STEAM TURBINES WITH A LOW-BOILING WORKING AGENT

    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.

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

    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.

  16. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    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.

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

    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.

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

    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. PMID:20666474

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

    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.)

  20. BORATING OF CARBON AND ALLOY STEEL IN BOILING LAYER

    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

  1. Corrosion fatigue behavior of zirconium in boiling nitric acid

    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)

  2. Calculation of boiling model change wave propagation rate

    Approximate analytical expression for the boiling mode change wave front rate on the rod and on the plate was obtained. The influence of the Thomson effect and of heater orientation in the gravitational field was taken into account. Paper shows satisfactory agreement of the obtained ratios with the experimental data

  3. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    D. S. Obukhov

    2014-01-01

    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.

  4. Pressure drop of subcooled flow boiling in narrow tube

    The pressure drop of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the conditions: tube inside diameter: 1 and 3mm, tube length: 10∼100mm, and water mass velocity: 7000∼20000kg/m2s. The friction pressure drop ratio of subcooled flow boiling to non-heating water flow was examined by increasing the heat flux. The ratio begins to increase at the heat flux proposed by the Saha-Zuber correlation that the bubble begins to detach for 3mm inside diameter tube, though the heat flux is higher than the Saha-Zuber heat flux for 1mm tube. The ratio was further compared with the Bergles-Dormer correlation. The two phase friction multiplier of subcooled flow boiling was examined assuming the Ahmad void fraction and applied to the Lockhart-Martinelli (L-M) correlation. The abnormarity of the subcooled flow boiling in the case of 1mm inside diameter tube was confirmed in these discussions. (author)

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

    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)

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

    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)

  7. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    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)....

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

    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

  9. Modeling acid-gas generation from boiling chloride brines

    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

  10. Modeling acid-gas generation from boiling chloride brines

    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

  11. Modeling acid-gas generation from boiling chloride brines

    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

  12. Phase relations and adiabats in boiling seafloor geothermal systems

    Bischoff, James L.; Pitzer, Kenneth S.

    1985-11-01

    Observations of large salinity variations and vent temperatures in the range of 380-400°C suggest that boiling or two-phase separation may be occurring in some seafloor geothermal systems. Consideration of flow rates and the relatively small differences in density between vapors and liquids at the supercritical pressures at depth in these systems suggests that boiling is occurring under closed-system conditions. Salinity and temperature of boiling vents can be used to estimate the pressure-temperature point in the subsurface at which liquid seawater first reached the two-phase boundary. Data are reviewed to construct phase diagrams of coexisting brines and vapors in the two-phase region at pressures corresponding to those of the seafloor geothermal systems. A method is developed for calculating the enthalpy and entropy of the coexisting mixtures, and results are used to construct adiabats from the seafloor to the P-T two-phase boundary. Results for seafloor vents discharging at 2300 m below sea level indicate that a 385°C vent is composed of a brine (7% NaCl equivalent) in equilibrium with a vapor (0.1% NaCl). Brine constitutes 45% by weight of the mixture, and the fluid first boiled at approximately 1 km below the seafloor at 415°C, 330 bar. A 400°C vent is primarily vapor (88 wt.%, 0.044% NaCl) with a small amount of brine (26% NaCl) and first boiled at 2.9 km below the seafloor at 500°C, 520 bar. These results show that adiabatic decompression in the two-phase region results in dramatic cooling of the fluid mixture when there is a large fraction of vapor.

  13. Bubble and boundary layer behaviour in subcooled flow boiling

    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)

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

    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.

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

    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...

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

    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)

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

    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)

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

    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

  19. An experimental study of flow boiling in a rectangular channel with offset strip fins

    An experimental study on saturated flow boiling heat transfer of R113 was performed in a vertical rectangular channel with offset strip fins. Two-phase pressure gradients and boiling heat transfer coefficients in an electrically heated test section were measured for the quality range of 0-0.6, mass flux range of 17-43 kg/m2 s and heat flux of 500-3000 W/m2. Two-phase frictional multiplier was determined as a function of Martinelli parameter. The two-phase forced convective component of the local boiling heat transfer coefficient was found to be well correlated with the Reynolds number factor. A superposition method for the flow boiling heat transfer coefficient that included the contribution of saturated nucleate boiling was verified also for flow boiling in a channel with offset strip fins. The predictions of local flow boiling heat transfer coefficients were found to be in good agreement with experimental data

  20. Control of the boiling crisis: analysis of a model system

    Controlling the transition between the low (nucleate) and high temperature (film) regimes of boiling is a serious challenge for a number of technological applications. Based on the theoretical analysis of a simplified reaction-diffusion model, it has recently been shown that the transition towards the dangerous situation where the high temperature phase tends to invade the whole system requires a higher power in a periodically spatially modulated system than in an homogeneous system. We show here that the transition mechanisms between the various boiling regimes depend on the ratio between the periodicity length along the wire and the characteristic thermal diffusion length. We analyse theoretically a simple experimental setup aimed at testing these ideas. The heater consists of a thin wire, with an applied electric current, with alternatively low resistance and high resistance sections. We determine the gain in stability for a set of realistic values of the parameters. (authors)

  1. Boiling flow simulation in Neptune-CFD and Fluent codes

    This paper presents simulations of the convective boiling flow performed with NEPTUNE-CFD and FLUENT codes. The DEBORA experiments carried out at CEA Grenoble were used as an experimental data set. In these experiments, freon R12 flows upwards inside a vertical pipe. Radial profiles of the flow variables are measured at the end of the heated section. Seven DEBORA cases were selected for simulation. NEPTUNE-CFD code was used without modifications because it contains all necessary models. In FLUENT, an important part of the models has been implemented by programming in User Defined Functions. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and mean bubble diameter at the end of the heated section shows that both codes can provide reasonable results in boiling conditions. The presented work was carried out within the 6. Framework EC NURESIM project. NEPTUNE-CFD code is implemented in the NURESIM platform. (authors)

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

    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)

  3. Acoustic measurement of boiling instabilities in a solar receiver

    Beattie, A. G.

    1980-11-01

    An acoustic technique was developed and used to search for boiling instabilities in the prototype receiver for the Barstow 10 MW Solar Thermal Pilot Plant. Instabilities, consisting of movements of the transition zone between regions of nucleate and film boiling, were observed. The periods of these fluctuations ranged between three and fifteen seconds with no indications of preferred frequencies. The peak to peak amplitudes of the fluctuations averaged 0.4 meters under steady state conditions at absorbed power levels between 2.0 and 3.2 MW. Transient fluctuations with amplitudes up to 2.0 meters were also seen. These transients usually lasted between 30 and 300 seconds. It was not possible to pinpoint the causes of these transients.

  4. Flow pattern characteristics of pool boiling in inclined confined spaces

    In this paper, visualization of bubble behavior and two-phase flow in inclined confined spaces are performed for near-saturated demineralized water at atmospheric pressure with gap sizes of 3 mm to 8 mm, and inclination angles of 0° to 300°. Based on the results, three boiling regimes are observed: isolated deformed bubbles, coalesced deformed bubbles and partial dry-out. A flow pattern map for confined pool boiling, based on the Bond number and a dimensionless number of heat flux, has been developed in order to determine the regimes. Objective criteria are proposed for the transitions between various regimes. It is shown that the transition criteria between isolated deformed bubbles and coalesced deformed bubbles is Q=0.16 and Q=0.55 between coalesced deformed bubbles and partial dry-out. It has to be pointed out that the gap structures and downward facing heating surfaces may be responsible for the special flow patterns. (authors)

  5. Critical heat flux maxima during boiling crisis on textured surfaces

    Dhillon, Navdeep Singh; Buongiorno, Jacopo; Varanasi, Kripa K.

    2015-09-01

    Enhancing the critical heat flux (CHF) of industrial boilers by surface texturing can lead to substantial energy savings and global reduction in greenhouse gas emissions, but fundamentally this phenomenon is not well understood. Prior studies on boiling crisis indicate that CHF monotonically increases with increasing texture density. Here we report on the existence of maxima in CHF enhancement at intermediate texture density using measurements on parametrically designed plain and nano-textured micropillar surfaces. Using high-speed optical and infrared imaging, we study the dynamics of dry spot heating and rewetting phenomena and reveal that the dry spot heating timescale is of the same order as that of the gravity and liquid imbibition-induced dry spot rewetting timescale. Based on these insights, we develop a coupled thermal-hydraulic model that relates CHF enhancement to rewetting of a hot dry spot on the boiling surface, thereby revealing the mechanism governing the hitherto unknown CHF enhancement maxima.

  6. Numerical modeling of boiling heat transfer in porous media

    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

  7. Consequences of select boiling waste scenarios for waste feed delivery

    The purpose of this calculation is to quantify thermal phenomena and predict consequences for select double-shell tank (DST) heat-up scenarios for waste feed delivery safety applications. This work extends previous tank bump analyses [Epstein, et al., 2000; and Epstein, et al., 2000a] by focusing on selected scenarios in which extended loss of cooling leads to waste boiling and aerosol generation. The HADCRT code, Version 1.3, is used to provide a best-estimate, coupled simulation of both thermal-hydraulic and aerosol transport phenomena. The physical model employed considers boiling aerosol generation, aerosol agglomeration and gravitational sedimentation in the tank headspace, aerosol removal FR-om steam condensation, and aerosol transport FR-om the headspace to the environment and through ventilation flow paths. Thus, the net result of the calculation is the time history of aerosol release to the immediate vicinity of an underground DST

  8. Vertical clog flow heat transfer with nucleate boiling

    This paper presents a model for slug flow heat transfer which is modified by incorporating the effect of nucleate boiling. This modification has made the slug flow model more useful for practical situations where presence of nucleate boiling is generally a norm rather than an exception. The model is further improved by taking into account the previously ignored effect of vapor bubbles present in the liquid slug. These modifications not only make the slug flow model more realistic but also improve its predictive capabilities. This improvement is demonstrated by comparing the predictions of the current model and the previous model with the published slug flow data. The mean deviation between the prediction and the measured slug flow data for water in the range of 20 to 100 kW/m2 heat flux is 5.87% for a total 171 data points

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

    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

  10. Saturation conditions in elongated single-cavity boiling water targets

    Steyn, G. F.; Vermeulen, C.

    2015-01-01

    Introduction It is shown that a very simple model reproduces the pressure versus beam current characteristics of elongated single-cavity boiling water targets for 18F production surprisingly well. By fitting the model calculations to measured data, values for a single free parameter, namely an overall heat-transfer coefficient, have been extracted for several IBA Nirta H218O targets. IBA recently released details on their new Nirta targets that have a conical shape, which constitutes an im...

  11. Superfluid helium boiling in porous structure under microgravity: model representation

    The results of model calculation of superfluid helium boiling under microgravity conditions are reported. The evolution of a vapour film on the cylindrical heater surface inside the porous thick walled structure is analyzed. The molecular-kinetic theory methods are used to describe heat and mass-transfer within helium interface. The equation of vapour-liquid interface motion is solved. The effect of experimental parameters on the vapour film properties is studied. The calculation data for microgravity and terrestrial conditions are compared

  12. Revisting the boiling of quark nuggets at nonzero chemical potential

    Li, Ang; Liu, Tong; Gubler, Philipp; Xu, Ren-Xin

    2013-01-01

    The boiling of possible quark nuggets during the quark-hadron phase transition of the Universe at nonzero chemical potential is revisited within the microscopic Brueckner-Hartree-Fock approach employed for the hadron phase, using two kinds of baryon interactions as fundamental inputs. To describe the deconfined phase of quark matter, we use a recently developed quark mass density-dependent model with a fully self-consistent thermodynamic treatment of confinement. We study the baryon number li...

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

    Orzechowski T.; Tyburczyk A.

    2014-01-01

    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 h...

  14. Calculation of limit cycle amplitudes in commercial boiling water reactors

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  15. Problem of Boil - off in LNG Supply Chain

    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...

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

    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

  17. Electrical control and enhancement of boiling heat transfer during quenching

    Shahriari, Arjang; Hermes, Mark; Bahadur, Vaibhav

    2016-02-01

    Heat transfer associated with boiling degrades at elevated temperatures due to the formation of an insulating vapor layer at the solid-liquid interface (Leidenfrost effect). Interfacial electrowetting (EW) fields can disrupt this vapor layer to promote liquid-surface wetting. We experimentally analyze EW-induced disruption of the vapor layer and measure the resulting enhanced cooling during the process of quenching. Imaging is employed to visualize the fluid-surface interactions and understand boiling patterns in the presence of an electrical voltage. It is seen that EW fields fundamentally change the boiling pattern, wherein a stable vapor layer is replaced by intermittent wetting of the surface. Heat conduction across the vapor gap is thus replaced with transient convection. This fundamental switch in the heat transfer mode significantly accelerates cooling during quenching. An order of magnitude increase in the cooling rate is observed, with the heat transfer seen approaching saturation at higher voltages. An analytical model is developed to extract voltage dependent heat transfer rates from the measured cooling curve. The results show that electric fields can alter and tune the traditional cooling curve. Overall, this study presents an ultralow power consumption concept to control the mechanical properties and metallurgy, by electrically tuning the cooling rate during quenching.

  18. Measurement of film dynamics in a boiling liquid film

    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)

  19. Film boiling on porous layered brass sphere during quenching

    Kang, Jun-young; Kim, Seol Ha; Jo, Hangjin; Lee, Gi Cheol; Kiyofumi, Moriyama; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Kim, Moo Hwan [KOREA Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    Fluid (liquid or gas) can afford to be permeable into porous layer on heat transfer surface and this phenomenon significantly affects phase-change heat transfer, especially boiling. The Corrosion Residual Unidentified Deposition (CRUD) which has generally micro-scaled pore geometry could have considered as porous layer and it was suggested that modification of heat transfer surface like CRUD can influence cooling rate during Loss-Of-Coolant Accident (LOCA) transient. Therefore, role of porous layer will be more emphasized at core-safety analysis, because, recently strategy of nuclear-fuel operation gradually becomes higher burn-up and longer cycle. As another aspect, study about film boiling has widely concerned due to its importance at core-coolability in LOCA, however, consideration of porous layer has relatively restricted because of difficulty of fabrication, excepting for horizontal surface. In this article, we briefly introduce experimental result of film boiling on porous layered surface during quenching. Laboratory-scaled quenching facility was applied and porous layer was fabricated by Electro-Chemical Deposition (ECD) method at spherical brass test section. We observed that the existence of porous layer on heat transfer surface considerable affected the cooling rate (t{sub cool,MPS}/t{sub cool,BBS}-12) during quenching in a saturated distilled water, therefore, it is expected that porous layer like CURD may have the potential able to affect LOCA transient.

  20. Film boiling characteristics of liquid metals in Leidenfrost phenomenon

    In a steel container filled with 1.0 bar argon gas, potassium or sodium droplets were put on a heated plate and observed by a camera and a VTR through viewports. In the case of potassium plate of 316-stainless steel and tantalum were used respectively for low and high temperatures and joule-heated by a DC power supply, while in the case of sodium a tantalum disk was induction-heated by a 20 kHz and 30 kW power supply. Stable film boiling was observed for a high temperature of the plate or disk, whereas an instantaneous breakup of droplet with extensive vaporization occurred for a low temperature. Whether or not a droplet did float on the plate was mainly governed by the interface temperature at the contact. The heat flux was estimated from volumetric reducing rate of droplet observed. For potassium the results in the film boiling region showed an appreciably good agreement with prediction based on the Bromley's expression combined with the theory of Baumeister et al. For sodium, the experimental results were slightly higher than the prediction. The minimum film boiling temperature and heat flux were roughly estimated to be about 13000C and 15W/cm2 for potassium and about 16000C and 45W/cm2 for sodium, respectively

  1. Phase field model for the study of boiling

    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)

  2. Flow boiling heat transfer in mini-channels

    In view of practical significance of a correlation of heat transfer coefficient in the aspect of such applications as engineering design and prediction, some efforts towards correlating flow boiling heat transfer coefficients for mini-channels have been made in this study. Based on analyses of existing experimental investigations of flow boiling, it was found that liquid-laminar and gas-turbulent flow is a common feature in many applications of mini-channels. Traditional heat transfer correlations for saturated flow boiling were developed for liquid-turbulent and gas-turbulent flow conditions and thus may not be suitable in principle to be used to predict heat transfer coefficients in mini-channels when flow conditions are liquid-laminar and gas-turbulent. By considering flow conditions (laminar or turbulent) in the Reynolds number factor F and single-phase heat transfer coefficient hsp, the Chen correlation has been modified to be used for four flow conditions such as liquid-laminar and gas-turbulent one often occurring in mini-channels. A comparison of the newly developed correlation with various existing data for mini-channels shows a satisfactory agreement. In addition, an extensive comparison of existing general correlations with databases for mini-channels has also been made. (author)

  3. Boiling Heat Transfer Experiments by using Transparent Heated Microtube

    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.

  4. Two-phase boundary layer prediction in upward boiling flow

    In the present work, the numerical modelling of the two-phase turbulent boundary layer in upward boiling flow was investigated. First, non-dimensional liquid velocity and temperature profiles in the two-phase boundary layer were validated on the one-dimensional section of a pipe with prescribed radial void fraction profiles. Simulations were performed on a fine grid with a commercial code CFX-5 using the k-ω turbulence model. A significant deviation of results from the analytical single-phase and two-phase wall functions from the literature was observed. Second, a wall boiling model in a vertical heated pipe was simulated (CFX-5) on the coarse grid. Here the prediction of the two-phase thermal boudary layer was compared to the experimental data, k-ω calculation on the fine grid and against the singlephase analytical wall function. Again a major deviation against single-phase temperature wall function was obtained. Presented analyses suggest that the existing analytical velocity and temperature wall functions cannot be valid for the boiling boundary layer with the high void fraction on the wall. (author)

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

    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)

  6. Intermittent phenomena in the boiling two-phase boundary layer

    In order to investigate statistical properties of temperature fluctuation in a boiling two-phase boundary layer the corresponding intermittency functions, which describe liquid, vapour and interface region at an individual fixed point, have been defined. In water boiling on a horizontal surface the temperature fluctuation was measured with a microthermocouple and the signal was processed through the digital computer with the detector function specified for liquid, vapor and interface region. The results obtained confirm that the temperature fluctuation in the boiling two-phase layer can be divided into three parts corresponding to individual regions and that its statistical distribution depends on the properties of respective systems. It has also been shown that the temperature fluctuation in the interface region is determinative and corresponds to the temperature changes in the liquid layer surrounding vapor bubble growth. Amplitude distribution in the liquid region changes its form with the distance from the wall as a result of the change in intensity of turbulence at different distances. The probability density distribution in the vapor region shows very small amplitude fluctuation and is almost constant for all distances. (author)

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

    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.

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

    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

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

    Kutnjak, Josip

    2013-06-27

    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

  10. Measurement of Key Pool Boiling Parameters in nanofluids for Nuclear Applications

    Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e., contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout.

  11. Measurement of key pool boiling parameters in nanofluids for nuclear applications

    Nanofluids, colloidal dispersions of nanoparticles in a base fluid such as water, can afford very significant Critical Heat Flux (CHF) enhancement. Such engineered fluids potentially could be employed in reactors as advanced coolants in safety systems with significant safety and economic advantages. However, a satisfactory explanation of the CHF enhancement mechanism in nanofluids is lacking. To close this gap, we have identified the important boiling parameters to be measured. These are the properties (e.g., density, viscosity, thermal conductivity, specific heat, vaporization enthalpy, surface tension), hydrodynamic parameters (i.e., bubble size, bubble velocity, departure frequency, hot/dry spot dynamics) and surface conditions (i.e., contact angle, nucleation site density). We have also deployed a pool boiling facility in which many such parameters can be measured. The facility is equipped with a thin indium-tin-oxide heater deposited over a sapphire substrate. An infra-red high-speed camera and an optical probe are used to measure the temperature distribution on the heater and the hydrodynamics above the heater, respectively. The first data generated with this facility already provide some clue on the CHF enhancement mechanism in nanofluids. Specifically, the progression to burnout in a pure fluid (ethanol in this case) is characterized by a smoothly-shaped and steadily-expanding hot spot. By contrast, in the ethanol-based nanofluid the hot spot pulsates and the progression to burnout lasts longer, although the nanofluid CHF is higher than the pure fluid CHF. The presence of a nanoparticle deposition layer on the heater surface seems to enhance wettability and aid hot spot dissipation, thus delaying burnout. (author)

  12. Study on pool-nucleate boiling heat transfer characteristics by using artificial cavities

    Pool boiling heat transfer experiments were performed by using the well-controlled/defined heat transfer surface for water. Uni-size and -shape artificial cavities were created on the mirror-finished silicon plate by utilizing the MEMS technology. Experimental results agreed well with what were predicted by the traditional boiling theory. The mirror finished surface showed only the tendency of natural circulation heat transfer. The artificial-cavity heat transfer surface followed the pool-nucleate boiling trend. The onset of the pool-nucleate boiling was well predicted by the traditional pool-nucleate boiling theory. These results indicated that the artificial cavities behave just like natural cavities. The results indicated the artificial cavities are quite useful and promising to examine the true features of complicated boiling that have been overshadowed by complicatedness. (authors)

  13. US licensing process and ABWR certification

    The Part 50 licensing process establishes a two-step licensing process whereby the U.S. Nuclear Regulatory Commission (NRC) authorizes construction through issuance of a construction permit and authorizes operation by issuance of an operating license. At each stage, the NRC staff conducts technical reviews, and there is potential for public hearings. The purpose of the Part 52 licensing process is to provide a regulatory framework that brings about early resolution of licensing issues in comparison with the Part 50 process. Because issues are not resolved early in the Part 50 licensing process, approval of an operating license is not assured until after a significant investment has been made in the plant. Part 52 increases the stability and certainty of the licensing process by providing for the early resolution of safety and environmental issues. The Part 52 licensing process features early site permits, design certification, and combined construction permits and operating licenses

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

    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.

  15. Upward Flow Boiling to DI-Water and Cuo Nanofluids Inside the Concentric Annuli

    N. Vaeli; M. M. Sarafraz; Peyghambarzadeh, S. M.; F Hormozi

    2015-01-01

    In this work, flow boiling heat transfer coefficients of deionized water and copper oxide water-based nanofluids at different operating conditions have been experimentally measured and compared. The liquid flowed in an annular space. According to the experiments, two distinguished heat transfer regions with two different mechanisms can be seen namely forced convective and nucleate boiling regions. Results demonstrated that with increasing the applied heat flux, flow boiling heat transfer coef...

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

    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.

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

    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

  18. Pressure drop across micro-pin heat sinks under boiling conditions

    Koşar, Ali; Kosar, Ali; Özdemir, Mehmed Rafet; Ozdemir, Mehmed Rafet; Keskinöz, Mehmet; Keskinoz, Mehmet

    2009-01-01

    Two-phase pressure drop was studied in four different micro pin fin heat sinks. Micro pin fin heat sinks used in the current studies were operated under boiling conditions using water and R-123 as working fluids. It was observed that once boiling was initiated severe temperature fluctuations and flow oscillations were recorded for three of the micro pin fin heat sinks, which was characterized as unstable boiling. Pressure drop signals were presented just before and after the unstable boili...

  19. Boiling heat transfer in thin liquid layers of mercury under magnetic field

    Experimental data are presented on the boiling heat transfer from a horizontal plane heater to liquid layers of mercury in the presence of a magnetic field of which direction is parallel to the direction of gravity. Increasing the magnetic flux density, the incipient boiling heat flux and burnout heat flux decrease in compare with those of no magnetic field. However, when the liquid layer is thin, the magnetic field affects them little. The visual study of mercury boiling experiment is also performed. (author)

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

    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)

  1. Influence of surface topography in the boiling mechanisms

    Highlights: • Pool boiling heat transfer. • Use of micro-textured surfaces to enhance heat transfer. • Importance of the bubble dynamics and of the interaction mechanisms in the overall heat transfer efficiency. • Effect of the micro-textures on bubble dynamics as a way to enhance pool boiling heat transfer. - Abstract: The present paper addresses the qualitative and quantitative analysis of the pool boiling heat transfer over micro-structured surfaces. The surfaces are made from silicon chips, in the context of pool boiling heat transfer enhancement of immersion liquid cooling schemes for electronic components. The first part of the analysis deals with the effect of the liquid properties. Then the effect of surface micro-structuring is discussed, covering different configurations, from cavities to pillars being the latter used to infer on the potential profit of a fin-like configuration. The use of rough surfaces to enhance pool boiling mainly stands on the arguments that the surface roughness will increase the liquid–solid contact area, thus enhancing the convection heat transfer coefficient and will promote the generation of nucleation sites. However, one should not disregard bubble dynamics. Indeed, the results show a strong effect of bubble dynamics and particularly of the interaction mechanisms in the overall cooling performance of the pair liquid–surface. The inaccurate control of these mechanisms leads to the formation of large bubbles and strong vertical and horizontal coalescence effects promote the very fast formation of a vapor blanket, which causes a steep decrease of the heat transfer coefficient. This effect can be strong enough to prevail over the benefit of increasing the contact area by roughening the surface. For the micro-patterns used in the present work, the results evidence that one can reasonably determine guiding pattern characteristics to evaluate the intensity of the interaction mechanisms and take out the most of the

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

    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)

  3. Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

    An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

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

    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)

  5. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author)

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

    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.

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

    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)

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

    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)

  9. Nucleate Boiling Heat Transfer of Nanofluids with Carbon Nanotubes on Plain and Low Fin Surfaces

    Nuclear power generation is being discussed in many countries as an alternative method to solving the world's energy crisis. For the safe operation of nuclear power plants, ways for increasing the critical heat flux (CHF) related to a loss of coolant accident are being investigated. In the event that the local heat flux exceeds the CHF, there is an abrupt shift in the boiling curve such that the nucleate boiling ceases and transition boiling and ultimately film boiling occur, finally resulting in a physical break down of the surface. Therefore, it is essential to maximize the CHF for the protection of nuclear power plants with maximum system performance. For the past decade, as a lot of research has been carried out for an improvement of the boiling heat transfer coefficients (HTCs) and CHF, new methods employing nano particles have been proposed. The objectives of this study are to measure the pool boiling HTCs of the water without and with carbon nanotubes (CNTs) on plain and low fin surfaces up to the CHF, and to analyze the effect of CNTs on both nucleate boiling HTCs and CHF. Pool boiling HTCs on all surfaces tested in water without and with CNTs increased as the heat flux increased, which is a typical trend in the pool boiling of pure fluids. For nanofluid with CNTs on low fin surfaces, the surface geometry and nano particles produced a double effect of increasing the CHFs

  10. Exploring the Limits of Boiling and Evaporative Heat Transfer Using Micro/Nano Structures

    Lu, Ming-Chang

    2010-01-01

    This dissertation presents a study exploring the limits of phase-change heat transfer with the aim of enhancing critical heat flux (CHF) in pool boiling and enhancing thermal conductance in heat pipes. The state-of-the-art values of the CHF in pool boiling and the thermal conductance in heat pipes are about two orders of magnitudes smaller than the limits predicted by kinetic theory. Consequently, there seems to be plenty of room for improvement. Pool boiling refers to boiling at a surface im...

  11. Integrated particle imaging velocimetry and infrared thermometry for high resolution measurement of subcooled nucleate pool boiling

    High-resolution data of nucleate pool boiling are important for the development of mechanistic numerical models based on computation fluid dynamics (CFD). This paper describes an innovative experimental facility that allows time- and space- resolved measurement of a series of important boiling-relevant quantities, including velocity and temperature distributions around individual bubbles, bubble shape, departure diameter and frequency, all of which can be used for validation of CFD-based models of boiling heat transfer. The facility comprises a temperature controlled boiling cell, with an indium tin oxide (ITO) heater and integrated high-frequency Particle Imaging Velocimetry (PIV) and Infrared Thermometry (IR). Some representative results are reported. (author)

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

    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)

  13. Pool boiling in microgravity with a single specie system

    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...

  14. Loss of coolant accident at boiling water reactors

    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)

  15. Safety systems and features of boiling and pressurized water reactors

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  16. Specialists' meeting on sodium boiling noise detection. Summary report

    The purpose of the meeting was to review and discuss methods available for detection of the initial stage of accidents in fast reactors, with most attention on reliable detection by acoustic techniques, which could provide a valuable addition to the safety protection. Results obtained from reactor experiments were also discussed and recommendations made for future developments. The meeting was divided into five technical sessions as follows: Signals from sodium boiling; Transmission of acoustic waves and background noise; Detection techniques; Reactor experiments; and Future requirements

  17. Flow pattern at critical condition in forced convection boiling

    An experimental investigation on flow pattern at critical condition (burnout) in forced convection boiling was carried out using R-113 as a working fluid. The test section was an internally heated vertical annular channel with a stainless-steel heater tube of 10 mm O. D. and a glass shroud of 22 mm I. D.. The flow pattern was identified by means of photographic observation and statistical nature of void fraction. Measurements were performed at the pressure 0.3 MPa, mass flux of 500 to 2000 kg/m2s, inlet subcooling of 0 to 58 K. (author)

  18. Contaminant Recovery during In-Situ Boiling in Rock

    Chen, F.; Liu, X.; Falta, R. W.; Murdoch, L. C.

    2009-12-01

    In-situ boiling may be an effective mechanism for removing contaminants from tight rock matrix where they would otherwise be all but inaccessible. Heating the matrix above the boiling temperature and then depressurizing will induce boiling that leads to large gas-phase pressure gradients and a steam stripping effect that can remove the contaminants from the matrix. Despite the promise of this process, it has not yet been demonstrated in the field or laboratory, and the controlling parameters and limits of the process are poorly understood. The objective of this project is to characterize mass transfer during boiling in saturated rock. We built an experimental apparatus to heat cores (5cmx30cm) of contaminated rock in a pressurized vessel. The core was sealed in Teflon tube with metal end caps and wrapped with a strip heater. Additional heaters were located in the end caps. Sensors were placed on the surface and embedded within the core to monitor the temperature. An insulation layer covered the strip heater to minimize the heat loss. A recent test was conducted using Berea sandstone (18 millidarcy) initially saturated with de-aired water and contaminated by injecting 200ml (about 2 pore volumes) containing 200mg/L of 1,2-dichloroethane (1,2-DCA), 10 mg/L of chlorobenzene (CB), and 195 mg/L sodium bromide (NaBr). The solution was circulated and both inlet and outlet concentrations were monitored. After the contaminant injection, both the inlet and outlet valves were closed and the core was heated at a constant power of 31.3 watts. Pressure and temperature increased for 3 hours until temperatures exceeded 100 C. A valve on the outlet tube was opened and steam flow started immediately and was routed through a condenser. Concentrations of chlorinated solvents in the outflow increased abruptly to between 6 and 10 times the input concentration. The concentrations decreased after a few 10s of ml were recovered, and at least 80 to 90 percent of the contaminant masses were

  19. Needs of nuclear data for advanced light water reactor

    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)

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

    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.

  1. Pool boiling performance of NovecTM 649 engineered fluid

    A new fluorinated ketone, C2F5C(O)CF(CF3)2, is currently being considered as an environmentally friendly alternative for power electronics cooling applications due to its high dielectric strength and low global warming potential (GWP). Sold commercially by the 3M Company as NovecTM 649 Engineered Fluid, C2F5C(O)CF(CF3)2 exhibits very low acute toxicity while maintaining long-term stability. To assess the general two-phase heat transfer performance of NovecTM 649, pool boiling tests were conducted by resistively heating a 0.01 in. diameter nickel wire at the fluid's atmospheric saturation temperature of 49 deg C. The nucleate boiling heat transfer coefficient and critical heat flux (CHF) obtained for the fluorinated ketone compare favorably with results obtained for FC-72, a fluorocarbon widely used for the direct cooling of electronic devices. Initial results indicate that NovecTM 649 may prove to be a viable alternative to FC-72 and other halo alkanes for the cooling of high power density electronic devices. (author)

  2. On-line monitoring of boiling crevice chemistry evolution

    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)

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

    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.

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

    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

  5. High Pressure Boiling Water Reactor HP-BWR

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  6. SWR 1000: the Boiling Water Reactor of the future

    Siemens Power Generation Group (KWU) is currently developing - on behalf of and in close cooperation with the German nuclear utilities and with support from various European partners - Germany's next generation of boiling water reactor. This innovative design concept marks a new era in the successful tradition of boiling water reactor technology and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared lo 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. In addition, a state-of-the-art materials concept featuring erosion-resistant materials and low-cobalt alloys as well as cobalt-free substitute materials ensures a low cumulative dose for operating and maintenance personnel and also minimizes radioactive waste. (author)

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

    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

  8. Experimental Research on Flash Boiling Spray of Dimethyl Ether

    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.

  9. Radiolysis effects in sub-cooled nucleate boiling

    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 H2O2. 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 H2 (STP) kg-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 H2 (STP) kg-1 range. (author)

  10. Modelling of boiling bubbly flows using a polydisperse approach

    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)

  11. Thermodynamic study on redox reactions of boiling nitric acid solutions

    In order to understand corrosion of metals in nitric acid solutions, it is necessary to know the generation mechanism of high equilibrium potential in the solutions, especially under boiling conditions. Existing nitrogen oxides in nitric acid solutions were first analyzed by Raman spectroscopy and then existing amount of nitrogen oxides were examined by thermodynamic calculation using the SOLGASMIX software. The Raman spectroscopic analysis showed that the existing amount of un-dissociated HNO3 increased with increasing nitric acid concentration and solution temperature. The existing amount of NO2 also increased by thermal decomposition. The thermodynamic calculation showed that the important nitrogen oxides in nitric acid solutions are HNO3, NO3-, HNO2, NO2, and NO. The equilibrium potential of nitric acid solution is, however, mainly decided by the HNO3/HNO2 equilibrium. The thermodynamic calculation also suggested that the increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition of nitrous acid on the surface and the continuous removal of decomposition product from the solutions by boiling babbles. (author)

  12. Drag reduction of flow boiling with polymer additives

    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. Numerical modelling of boiling heat transfer in microchannels

    In this paper we report the results of our modelling studies on two-phase forced convection in microchannels using water as the fluid medium. The study incorporates the effects of fluid flow rate, power input and channel geometry on the flow resistance and heat transfer from these microchannels. Two separate numerical models have been developed assuming homogeneous and annular flow boiling. Traditional assumptions like negligible single-phase pressure drop or fixed inlet pressure have been relaxed in the models making analysis more complex. The governing equations have been solved from the grass-root level to predict the boiling front, pressure drop and thermal resistance as functions of exit pressure and heat input. The results of both the models are compared to each other and with available experimental data. It is seen that the annular flow model typically predicts higher pressure drop compared to the homogeneous model. Finally, the model has also been extended to study the effects of non-uniform heat input along the flow direction. The results show that the non-uniform power map can have a very strong effect on the overall fluid dynamics and heat transfer

  14. Radiolysis effects in sub-cooled nucleate boiling

    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)

  15. Subcooled film boiling heat transfer on a high temperature sphere in very dilute Al{sub 2}O{sub 3} nano-fluids

    Hyun Sun Park; Dereje Shiferaw; Bal Raj Sehgal [Royal Institute of Technology, Department of Energy Technology, Division of Nuclear Power Safety, Drotning Kristinas Vag 33A, SE-10044, Stockholm (Sweden)

    2005-07-01

    Full text of publication follows: nano-fluids, or conventional liquids, e.g., water, with small concentration of nano-particles uniformly suspended, have attracted attention as a new heat transport medium with enhanced thermo-physical properties. Up to the present, only exploratory experiments on nano-fluids have been reported. Das et al (Int. J. Heat Mass Transfer 43, pp 3701-3707, 2003) conducted boiling experiments with water containing 38 nm Al{sub 2}O{sub 3} nano-particles. They observed deterioration in the nucleate boiling heat transfer due to the deposition of nano-particles. Boiling experiments conducted by Vassallo et al (Int. J. Heat Mass Transfer 47, pp 407-411, 2004) using silica nano-fluid using 0.4 mm diameter NiCr wire showed three times higher critical heat flux (CHF) and the wire traversed the film boiling region before it failed. Another independent experiment performed on 1 cm{sup 2} square plate with a very low concentration of nano-particles ranging from 0.01 to 0.05 g/liter and at under pressure (2.89 psia), nano-fluids resulted in drastic 2{approx}3 times enhancement of the CHF (You and Kim, Appl. Phys. Lett. 83. No 16, 2003). However in all the aforementioned studies no appropriate explanation of the CHF enhancement has been advanced. The measured 2-3 times higher critical heat flux for very dilute nano-fluids may have high significance if such nano-fluids could be employed in heat transport systems. Recently, we investigated the effect of nano-particles on film boiling, which governs heat transfer during accident conditions in a reactor plant, e.g., in coolability of a degraded core, or a particulate debris bed or a core melt, and in steam explosions. Our previous experiments performed on film boiling in nano-fluids having larger concentrations of 5, 10, and 20 g/liter than those in You's experiments showed that the nano-fluids lower the film boiling temperature, decrease the film boiling heat transfer and provide a much thicker and

  16. Experimental investigation on critical heat flux and transition boiling of water flow under increased pressure

    In connection with reactor safety problems (LOCA) a measuring technique has been developed which enables, within the parameter range of medium pressure (0.11 MPa - 1.20 MPa) and low mass flow densities (10 kg/m2s - 500 kg/m2s), exact experimental investigations of critical heat flux and transition boiling of water under quasi-stationary conditions. The system consists of a vertical, temperature-controlled short test section with water flowing upwards inside; an experimental loop controllable to a large extent; a quick automatic data acquisition, and numeric evaluation procedures. Quasi-stationary measured boiling curves, from nucleate boiling to film boiling (circa 450deg C), demonstrate the importance of pressure, mass flow density, and inlet subcooling, the boiling pressure being the most important parameter. The linear course of the boiling curves during transition boiling is remarkable. A frequently suspected hysteresis of the boiling curve could not be detected. The influence of surface effects (contact angle) clearly decreases with increasing pressure. For the empirical correlation of the measured data by means of indices, a statement was chosen that normalizes the heat flux density of transition boiling to the maximum heat flux density at the beginning of the post-CHF range. As a result, the experimental data obtained, and the correlation developed from them, show a better heat transfer in transition boiling than conservatively assumed in general in literature. The temperature-controlled measurements of complete boiling curves supply data for critical heat flow density and the corresponding wall overheating. A comparison with the uncontrolled operation of the test section shows differences of 5-6% only within the range of measurement accuracy of such experiments. (orig.) With 36 figs., 5 tabs

  17. Results of boil-off experiment QUENCH-11

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor (LWR). The typical QUENCH test bundle consists of 21 fuel rod simulators with a total length of approximately 2.5 m. Boil-off experiment QUENCH-11 was performed at Forschungszentrum Karlsruhe (Karlsruhe Research Center) on 8 December, 2005 as the second of two experiments in the frame of the EC-supported LACOMERA program. The experiment focused on studying bundle behavior during boil-off and subsequent quenching at a small water injection rate. It was proposed by INRNE Sofia (Bulgarian Academy of Sciences) and defined together with the Karlsruhe Research Center. The analytical support by using the SCDAP/RELAP5 mod 3.2.irs and ASTEC codes for the preparation of the entire test was essential in conducting the test. A steady boil-off and a corresponding top-down uncovery of the test bundle was achieved by applying power from an electric auxiliary heater at the bundle bottom in addition to the electric bundle power. An additional outer heating system compensated heat losses that would lead to a reduction in boiling off the covered part of the bundle. When the water level had fallen to -70 mm elevation water was injected into the lower plenum at a rate of ca. 1 g/s enabling a nearly stable water level and an extension of the boil-off phase. Quenching of the bundle was performed at a maximum measured bundle temperature of 2040 K with a rather low mass flow rate of water, i.e. 18 (17+1) g/s, compared to the standard water injection rate of approx. 50 g/s. The conditions led to an enhanced cladding and shroud oxidation, quite similar to standard conditions of forced-convection steam flow. In the upper part the test bundle was significantly degraded by oxidation and melt formation. The total generation of hydrogen measured by the mass spectrometer was 141 g, of which 132 g, i.e. more than 90 % of the total, was produced during

  18. High-speed infrared thermography for the measurement of microscopic boiling parameters on micro- and nano-structured surfaces

    Micro- and nano-scale structures on boiling surfaces can enhance nucleate boiling heat transfer coefficient (HTC) and critical heat flux (CHF). A few studies were conducted to explain the enhancements of HTC and CHF using the microscopic boiling parameters. Quantitative measurements of microscopic boiling parameters are needed to understand the physical mechanism of the boiling heat transfer augmentation on structured surfaces. However, there is no existing experimental techniques to conveniently measure the boiling parameters on the structured surfaces because of the small (boiling on micro- and nano-structured surfaces. The visualization results are analyzed to obtain the microscopic boiling parameters. Finally, quantitative microscopic boiling parameters are used to interpret the enhancement of HTC and CHF. In this study, liquid-vapor phase distributions of each surface were clearly visualized by IR thermography during the nucleate boiling phenomena. From the visualization results, following microscopic boiling parameters were quantitatively measured by image processing. - Number density of dry patch, NDP IR thermography technique was demonstrated by nucleate pool boiling experiments with M- and N surfaces. The enhancement of HTC and CHF could be explained by microscopic boiling parameters

  19. On-site staffing requirements for a simplified boiling water reactor (SBWR)

    In 1992 the total generating costs were estimated by EPRI for a baseload, nth-of-a-kind advanced reactor with the following cost distribution: capital cost 62%, operation and maintenance (O and M) cost 20%, fuel cost 16%, and decommissioning cost 2%. Thus the O and M cost is a significant component of the total cost of electricity, second only to the capital cost. The O and M cost in turn can be split into: cost for on-site staff, maintenance materials, supplies and expenses, off-site technical support, regulatory fees, insurance premiums and administration. The costs for on-site staff is about 30% of the total O and M cost. In 1992, the US Council for Energy Awareness (USCEA) estimated the on-site staffing for a typical 600 MWe advanced reactor to be about 330 with 25 (full time equivalent, FTE) contractors. This estimate was reevaluated by EPRI, and the staffing was modified based on a reengineering of the organizational structure that eliminated unnecessary layers of vertical management. As a result of this review, the on-site staffing was decreased to 259 with 25 (FTE) contractors, for a total of 284 people. The Dodewaard power plant (GKN) in The Netherlands is a 60 MWe facility with a natural circulating reactor. Since the 600 MWe Simplified Boiling Water Reactor (SBWR), an advanced reactor, also utilizes a natural circulating reactor with other passive safety features it was desired to extrapolate the GKN staffing to the SBWR. Also, some of the European O and M practices that utilize fewer skilled labor are reflected. This paper provides the results of the comparison between the EPRI recommendations and the staffing based on GKN experience

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

    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.

  1. Crisis behaviour of the reactive recoil of a water jets under conditions of explosive boiling

    One presents the measurement results of the reactive force of boiling up water jet flowing through short channel into the atmosphere depending on superheating degree and at various evaporation mechanisms. The intensive fluctuation evaporation of water (explosive boiling) and presence of a plane perpendicular to the channel axis are shown to result in abrupt reduction of the reactive recoil value

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

    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…

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

    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'…

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

    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

  5. 46 CFR 154.706 - Cargo boil-off as fuel: Fuel lines.

    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...

  6. IAEA/IWGFR benchmark tests on sodium boiling noise detection. Part 1

    The present paper deals with investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. (author). 2 refs, 21 figs, 2 tabs

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

    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.

  8. Experimental study on augmentation of nucleate boiling heat transfer on nano porous surfaces

    Park, Young Jae; Kim, Hyung Dae [Kyung Hee Univ., Seoul (Korea, Republic of)

    2012-10-15

    Nucleate boiling broadly occurs in thermal hydraulic and safety systems of nuclear power plant (NPP). Heat transfer performance of nucleate boiling is closely related to efficiency and safety of NPPs. Hence, there have been numerous researches to effectively enhance nucleate boiling heat transfer performance. A number of recent studies have reported significant enhancements in nucleate boiling heat transfer coefficient (NBHTC) and critical heat flux (CHF) by fabricating nano/microscale structures on a boiling surface. Wei et al. showed that both NBHTC and CHF can be significantly enhanced with micro pin finned structures. They explained enhancement of NBHTC and CHF that occurred by increase in effective heat transfer area due to micro pin finned structures. Ahn et al. reported 100% enhancement in CHF on a boiling surface with nano/micro hybrid structures. They analyzed CHF enhancement that was caused by improvement of surface wettability on Nano/micro hybrid structures. In this study, an ordered nano porous surface was prepared using anodized aluminum oxide (AAO) technique and nucleate boiling heat transfer performance was examined in a pool with FC 72. Furthermore, the pool boiling result on the nano porous surface was interpreted based on heterogeneous bubble nucleation theory from a cavity.

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

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  10. 46 CFR 154.709 - Cargo boil-off as fuel: Gas detection equipment.

    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...

  11. 46 CFR 154.708 - Cargo boil-off as fuel: Valves.

    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...

  12. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has been... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section...

  13. 46 CFR 154.707 - Cargo boil-off as fuel: Ventilation.

    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)...

  14. Advanced robotic remote handling system for reactor dismantlement

    An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

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

    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.

  16. Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

    An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (orig.)

  17. High conversion pressurized water reactor with boiling channels

    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

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

    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

  19. High conversion pressurized water reactor with boiling channels

    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

  20. Early detection of coolant boiling in research reactors with MTR-type fuel

    Kozma, R.; Turkcan, E.; Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs.

  1. Early detection of coolant boiling in research reactors with MTR-type fuel

    In this paper, a reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University Delft, The Netherlands

  2. Early detection of coolant boiling in research reactors with MTR-type fuel

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

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

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (cylindrical boiling) facility, a reactor-scale facility which has a tank-within-a-tank design simulating the reactor vessel and the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm2 across the vessel bottom are performed. Boiling outside the reactor vessel is found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that the flooded cavity in a passive PWR like the AP-600 should be capable of cooling the RPV in the central region of the lower head. 9 figs., 1 tab., 14 refs

  4. Simulation of boiling flow experiments close to CHF with the NEPTUNE-CFD code

    A three-dimensional two-fluid code NEPTUNECFD has been validated against the ASU (Arizona State University) [1] and DEBORA [2, 3] boiling flow experiments. Nucleate boiling processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function is implemented in the NEPTUNECFD V1.0.6 code to improve the prediction of flow parameters in the boiling boundary layer. The capability of the code to predict boiling flow regime close to critical heat flux (CHF) conditions has been assessed on selected DEBORA experiments. It was shown that the code is able to predict wall temperature excursion and a sharp void fraction increase near the heated wall, which are characteristic phenomena for CHF conditions. (author)

  5. In-air PIXE for analyzing heavy metals in water boiled in pans

    The release rates of heavy metals from pans were measured for boiling water as well as for an acidic solution prior to an investigation on the release or sorption of trace elements due to cooking of food by boiling. The boiled samples were condensed and analyzed by means of in-air PIXE. The release of heavy metals was measured for five kinds of pans. For all pans the release rates were considerably more increased by boiling of a 5% solution of acetic acid. Furthermore it was found that by using the alumina coated aluminum pan (alumina pan) the respective release rates of Fe, Cu and Zn were all less than 50 μg per 100 cm2 of the pan surface dipped in the solution, and that monitoring of the contents of aluminum in the boiled solution enabled the estimation of the contribution of metal elements from the pan wall. (orig.)

  6. Pool boiling heat transfer on heterogeneous wetting surface with hydrophobic dots

    The boiling heat transfer mechanism of pool boiling is fundamental phenomena for understanding the phase change nature. Of many surface characteristics, the effects of wettability of heating surface is focused on as the dominant parameter for bubble dynamics and boiling heat transfer. In this study, highly controlled heating surfaces via MEMs technique were used for understanding the boiling heat transfer of heterogeneous wetting surfaces mixed by hydrophobic dots and a hydrophilic substrate. The diameter of hydrophobic dots and area ratio of phobic dots to heating area were regulated. The range of phobic dot diameter and area ratio were 50∼1000μm and 18.33∼54.3%, respectively. The performance of boiling heat transfer of each surface were evaluated by comparing with a wholly hydrophilic surface. It will contribute to understand the mechanism and criterion of enhanced heating surface condition by modified surface treatment procedure

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

    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.

  8. Acoustic measurements of the boiling stability tests on THORS sodium loop

    Acoustic data of boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility were obtained using three sodium-immersible high temperature microphones. The data was analyzed in both the time and frequency domains and provides the following information: (1) the acoustic signal due to sodium boiling was clearly observed; (2) the signal level and the repetition rate of boiling pulses are directly proportional to the applied heat flux; (3) a typical boiling pulse consists of a high frequency signal due mainly to the bubble collapses and a low frequency void oscillation; (4) the frequency spectra of the boiling and background pulses can be mostly assigned to various acoustic resonance frequencies of the THORS loop

  9. Modelling of CRUD growth phenomena on PWR fuel rods under nucleate boiling conditions

    PWR primary circuit materials undergo general corrosion leading to a release of metallic element release and subsequent process of particle deposition and ion precipitation on the primary circuit surfaces. The species accumulated on fuel rods are activated by neutron flux. Consequently, crud erosion and dissolution induce primary coolant contamination. In French PWRs, 58Co volume activity is generally low and almost constant (< 30 MBq.m-3) throughout an ordinary operating cycle. In some specific cases, a significant increase in volume activity is observed after the middle of a cycle (100-1000 MBq.m-3 for 58Co) when conditions for nucleate boiling are locally reached in certain fuel assemblies. Indeed, it is well known that nucleate boiling intensifies the deposition process. The thickness of the crud layer can reach some micrometers in non-boiling areas, whereas it can reach 100 micrometers in boiling areas. Crud growth in boiling conditions can be related to three phenomena: bubble growth induces deposition process (called boiling deposition), bubbles induce concentration increase at crud-coolant interface (called enrichment and modelled by the enrichment factor, the ratio between the wall concentration and the bulk concentration) and vaporisation induces concentration increase inside the crud. A literature review on the modelling of these phenomena and on the crud structure in nucleate boiling conditions has been carried out. The OSCAR [1] calculation code developed by the CEA to predict surface and volume activities in a single phase PWR primary circuit was chosen as a basis for present study. Ability to describe local nucleate boiling conditions was added to this code leading to realistic modelling of subsequent volume activity increase. In this article, we present the results obtained using a modified version of the OSCAR PC V1.2 calculation code including: - A double phase thermal-hydraulic module, - A model of boiling crud growth, able to calculate inner

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

    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)

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

    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.

  12. Channel-type nuclear reactor with a boiling coolant

    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

  13. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using ''representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs

  14. A stochastic study of noise in boiling reactors

    A stochastic point model is considered of a boiling reactor, involving the population of neutrons, the population of delayed neutron precursors, fuel temperature, and the number of bubbles in the coolant as random variables. Whereas the first two variables are related to capture, fission and delayed neutron processes, the other two take into account heat transfer between fuel and coolant and changes in coolant density. The variations in the fuel temperature and the coolant density generate reactivity feedbacks which affects the neutron power spectral density; analysis of the shape of this spectral density is expected to give information on the value of reactor parameters such as, for example, the void coefficient. (U.K.)

  15. Dynamic simulation of a boiling water nuclear reactor

    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)

  16. Theoretical prediction method of subcooled flow boiling CHF

    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)

  17. Interactions between bubble formation and heating surface in nucleate boiling

    Luke, Andrea [Leibniz University, Hannover (Denmark). Inst. of Thermodynamics], e-mail: ift@ift.uni-hannover.de

    2009-07-01

    The heat transfer and bubble formation is investigated in pool boiling of propane. Size distributions of active nucleation sites on single horizontal copper and steel tubes with different diameter and surface finishes have been calculated from heat transfer measurements over wide ranges of heat flux and selected pressure. The model assumptions of Luke and Gorenflo for the heat transfer near growing and departing bubbles, which were applied in the calculations, have been slightly modified and the calculated results have been compared to experimental investigations by high speed video techniques. The calculated number of active sites shows a good coincidence for the tube with smaller diameter, while the results for the tube with larger diameter describe the same relative increase of the active sites. The comparison of the cumulative size distribution of the active and potential nucleation sites demonstrates the same slope of the curve and that the critical radius of a stable bubble nuclei is smaller than the average cavity size. (author)

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

    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)

  19. Effects of Outflow Area on Pool Boiling in Vertical Annulus

    To identify the effects of an outflow area on pool boiling heat transfer in a vertical annulus, three different flow recreates were studied experimentally. For the test, a heated tube of smooth stainless steel and water at atmospheric pressure were used. Both annuli with open and closed bottoms were considered. To validate the effects of the outflow area on the heat transfer, the results of the annulus with the reactors were compared with the data for the plain annulus without the reactors. The reduction of the outflow area ultimately results in a decrease in the heat transfer. As the outflow area is very small, a slight increase in heat transfer is also observed. The major cause of this tendency is explained as the difference in the intensity of liquid agitation cause by the movement of coalesced bubbles. It is identified that the convective flow, pulsating flow, and evaporative mechanism are considered as the important mechanisms

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

    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.