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Sample records for boiling reactor experiment 5

  1. Mark I 1/5-sale boiling water reactor pressure suppression experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 1.3.1, 1.4, 1.5, and 1.6 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  2. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    International Nuclear Information System (INIS)

    Lai, W.; McCauley, E.W.

    1978-01-01

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

  3. Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lai, W.; McCauley, E.W.

    1978-01-04

    Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

  4. Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, D.; Lord, D.

    1978-03-16

    The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs.

  5. Operating experience with the multienrichment initial core of the boiling water reactor Kashiwazaki-Kariwa Unit 5

    International Nuclear Information System (INIS)

    Mochida, Takaaki; Nakamura, Mitsunari; Yamashita, Junichi; Maruyama, Hiromi; Muto, Sakae; Kasai, Shigeru

    1996-01-01

    The multienrichment boiling water reactor (BWR) initial core design was first applied to the Kashiwazaki-Kariwa Nuclear Power Station Unit 5 [1100-MW (electric) BWR] in Japan. This core is designed to improve fuel discharge exposure, capacity factors, and operability. The design study shows that three types of fuel bundles with different enrichments are suitable to achieve the design targets. Three bundle enrichments are selected to simulate each of the following: fresh bundles, once-burned bundles, and twice-burned bundles in the reload core. Although the heterogeneity of multienrichment design increases the complexity of the design analysis, both the initial criticality test and the moderator temperature coefficient measurement showed good agreement with the prediction. Subsequent full-power operation verified the expected core performance. Average discharge exposure for the total initial fuel is ∼10% larger than that for the conventional single-enrichment BWR initial fuel with reinsertion of discharged fuel at the end of the first cycle. These experiences verified the effectiveness of a multienrichment initial core for the improvement of fuel utilization, capacity factors, and operability

  6. Mark I 1/5-scale boiling water reactor pressure suppression experiment quick-look report

    International Nuclear Information System (INIS)

    McCauley, E.W.; Pitts, J.H.

    1977-01-01

    The tests conducted on the 1 / 5 -scale BWR Mark I pressure suppression test facility simulate the three-dimensional transient conditions that are encountered in a wetwell pressure suppression system during a hypothetical loss-of-coolant accident (LOCA). Specifically, the nitrogen (N2)-driven air clearing phase tests discussed here were performed to obtain the air/water-induced dynamic vertical load function and to determine the response of a 90 0 sector of a 360 0 torus structure

  7. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  8. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    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)

  9. Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

  10. A boiling-water reactor concept for low radiation exposure based on operating experience

    International Nuclear Information System (INIS)

    Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

    1983-01-01

    A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

  11. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1977-01-01

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  12. Gamma heated subassembly for sodium boiling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed.

  13. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    International Nuclear Information System (INIS)

    Huang Zhengyu; Edwards, Robert M.

    2003-01-01

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified

  14. Boiling-water reactor safety studies

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The Nuclear Regulatory Commission has funded LLL to study the pressure-suppression containment system of the Mark I class of boiling-water reactors (BWR). In particular, LLL is investigating how this containment system responds to a loss-of-coolant accident (LOCA), a design basis for light-water nuclear reactors. Part of this work is being carried out on the Laboratory's 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peachbottom 2 nuclear power plant. LLL is also conducting computer analyses of the way wall flexibility affects LOCA-induced loads in the containment system and of the safety margins in the containment structure. Results from these studies will help the NRC to review future BWR designs and may lead to decisions affecting the continued operation of many existing BWR power plants in the United States

  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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

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

  16. Advanced boiling water reactor (ABWR). Design, construction, operation and maintenance experience

    International Nuclear Information System (INIS)

    Idesawa, M.

    1998-01-01

    The ABWR has experienced all phases of design, construction, operation and maintenance at Kashiwazaki-Kariwa Nuclear Power Station Units No.6 and 7 and confirmed that originally intended development targets have been achieved with highly satisfactory results. This is the fruit of a project that collected wisdom from various sources under a international cooperative organization, with Tokyo Electric Power Company taking the leading role from the onset. These two units have not only demonstrated that ABWRs have superior performance as the first standard units of advanced light water reactor but also aroused a hope for the big potential advantages that ABWRs can provide us. The ABWR has already been awarded a U.S. standard license for having proved that it can comply with the requirements of international regulatory systems with an ample margin. There are also many construction programs with ABWRs progressing both domestically and abroad, suggesting that it has won recognition as an international standard plant. We will do our utmost to perfect the operation and maintenance records of Kashiwazaki-Kariwa Units No.6 and 7, which is the top runner among ABWRs, and to make known the superiority of this reactor to the world. (J.P.N.)

  17. Compact containment boiling water reactor (CCR)

    International Nuclear Information System (INIS)

    2006-01-01

    The compact containment boiling water reactor (CCR) is a modular boiling water reactor (BWR) designed by the Toshiba Corporation with the support of the Japan Atomic Power Company (JAPC). The current CCR design falls into the category of innovative small and medium size reactors, featuring 300MW electrical output per module. In Japan, increases in nuclear plant unit capacity have been promoted to take advantage of the economies of scale while further enhancing safety and reliability. As a result, more than 50 nuclear units are playing an important role in the domestic electric power generation. The next generation reactor with a 1700 MW(e) capacity is currently under development [IX-1, IX-2]. However, the future of nuclear power generation looks uncertain because of increasing competition with other power sources [IX-3] in the deregulated market, in spite of the general recognition that nuclear power is attractive from the viewpoint of energy security and environmental protection. Furthermore, factors such as stagnant growth in recent electricity demand, limitations in grid capacity and limited initial investment to avoid risk, will not favour large plant outputs. Nuclear plants are required that can easily be adopted in any country to globalize nuclear power generation for the mitigation of greenhouse effects. In the 1980s, the Toshiba Corporation has carried out R and D for BWRs with natural circulation and passive safety features. These R and D included tests and analysis of passive containment cooling systems (PCCS), isolation condensers (IC) and gravity driven cooling systems (GDCS). The results obtained through these tests have been used in the design of a simplified boiling water reactor (SBWR). Based on these activities, the design of a simplified BWR with a long operating cycle (LSBWR) design has been under development since the mid 1990s. The concept of the LSBWR is to provide flexibility to meet site conditions and electricity demands, to mitigate

  18. PSEPLOT: a controller for plotting data from the Mark I Boiling Water Reactor Pressure Suppression Experiment

    International Nuclear Information System (INIS)

    Holman, G.S.

    1978-01-01

    PSEPLOT is a computer routine that was developed for the Lawrence Livermore Laboratory Octopus computer system to generate several thousand plots of engineering data in a consistent format for referencing and comparison. The time-dependent engineering data were recorded during each of 25 tests of the Mark I Pressure Suppression Experiment (PSE). Although PSEPLOT is restricted to PSE, its concept is applicable to any similar data management task

  19. Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

    International Nuclear Information System (INIS)

    Beare, A.N.; Dorris, R.E.; Gray, L.H.

    1984-12-01

    This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

  20. Effects of Nanofluid for In-Vessel Retention External Reactor Vessel Cooling on Critical Heat Flux using Pool Boiling Experiments

    International Nuclear Information System (INIS)

    Park, Sung Dae; Kang, Sarah; Lee, Seung Won; Bang, In Cheol

    2011-01-01

    In-vessel retention (IVR) is one of the severe accident management (SAM) strategies that are used in some nuclear power plants: AP600, AP1000, Loviisa and APR1400. One way of IVR is the method of external reactor vessel cooling (ERVC). When core melts and deposits on the bottom of reactor vessel, ERVC is starting to flood the reactor cavity to remove the decay heat through the wall of the reactor vessel. This process can improve the plant economics by reducing regulatory requirements. And increased safety margin leads to gain public acceptance. In this system, the heat removal is restricted by thermal limit called by critical heat flux (CHF). Besides, as advanced light water reactors such as South Korea's APR-1400, thermal safety margin is deceased. So, it is essential to get more safety margin. There are some approaches to enhance the ERVC: using the coating on the vessel outer surface, increasing the reactor cavity flood level, streamlining the gap between the vessel and the vessel insulation. Many investigations have been performed to evaluate the coolability of IVR In this paper, we firstly investigated the coating effects in the critical heat flux among the above mentioned approach methods. During the boiling phenomenon, a thin layer was formed on the heater surface in the nanofluid. This coating mechanism is well known theoretically. Nanofluids are colloidal dispersions of nanoparticles in traditional heat transfer fluids. One of the most interesting characteristics of nanofluids is their capability to enhance the critical heat flux (CHF) significantly. Nanofluid is made by typical particle materials. Materials of nanoparticles include metals (e.g., silver, copper, gold), metal oxides (e.g., titania, alumina, silica, zirconia), carbon allotrope (e.g., carbon nanotube, graphite). We selected the grapheneoxide nanofluid which is a kind of carbon allotrope. Graphene-oxide is attractive material with the high thermal conductivity and stable dispersion ability in

  1. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  2. NUCLEAR SUPERHEATER FOR BOILING WATER REACTOR

    Science.gov (United States)

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

    1962-05-15

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

  3. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  4. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    Rasmusson, U.

    1983-04-01

    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)

  5. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  6. Computer simulations of a 1/5-scale experiment of a Mark I boiler water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1978-01-01

    The CHAMP computer code was employed to simulate a plane-geometry cross section of a Mark I boiling water reactor toroidal pressure suppression system air discharge experiment under hypothetical loss-of-coolant accident conditions. The experiments were performed at the Lawrence Livermore Laboratory on a 1 / 5 -scale model of the Peach Bottom Nuclear Power Plant

  7. Summary of the OECD/NRC Boiling Water Reactor Turbine Trip Benchmark - Fifth Workshop (BWR-TT5)

    International Nuclear Information System (INIS)

    2003-01-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. The purpose of this fifth workshop was to discuss the results from Phase III (best

  8. Progress on technology of boiling water reactor

    International Nuclear Information System (INIS)

    Ogawa, Nagao

    1975-01-01

    Progress has been made on the technology of boiling water reactors since the successful operation of Dresden BWR No.1. The technical advancement of BWRs has continued with the adoption of many kinds of proven techniques until the present stage. The advancement was made in the following items; improvement of core fuel, increase of plant power output, adoption of jet pump and moisture separator, improvement of containment and other items. Recently the technology of BWRs was reviewed from the point of nuclear plant safety and reliability and some new techniques are now under examination in order to apply to BWR plants. These items are as follows; improvement of core fuel assembly (adoption of 8x8 array fuel assembly), improvement of reactor recirculating system (flow control valve and jet pump), improvement of emergency core cooling system, revised control system, radioactive waste disposal system and adoption of standard design of BWR plants. These technical trend will produce more reliable and safer BWR plants. (Iwase, T.)

  9. Some fundamental aspects of boiling in nuclear reactors

    International Nuclear Information System (INIS)

    Mondin, H.; Lavigne, P.; Semeria, R.

    1964-01-01

    The main results obtained at Grenoble during the last four years in the field of boiling mechanisms and related phenomena in nuclear reactors are reported. 1 - Observation Of Boiling: By the use of photography and ultrafast cinematography (8000 frames per second maximum), boiling in a vessel or a tube was observed up to 140 kg/cm 2 . The populations of bubble-generating seeds (sites) were counted, and a correlation established giving their number per unit of surface area as a function of the thermal flux and the pressure. The diameter of the bubbles breaking of from the wall was studied up to 140 kg/cm 2 : three types of bubble have been shown to exist: - those in equilibrium, their diameter following the formula of Fritz and Ende, - bubbles found by boiling, the diameters of which decrease rapidly with the pressure (1/100 mm to 140 kg/cm 2 ), - the coalescences which appear in saturated liquid above 15 W/cm 2 , their proportion being independent of the pressure. Strioscopic observations were made of the movements of the thermal film associated with the generation of the seeds, at the initiation and condensation of the bubbles, the mechanisms responsible for the highly efficient heat transfer could thus be defined. 2 - Pressure Losses In Two-Phase Flow: A physical model of the continuous variation of the free space content in a boiling channel has been proposed by means of which the pressure losses can be calculated without invoking a break in the coefficient of friction when free boiling begins. Agreement between theory and experiment is satisfactory. The various forms which total pressure loss in a boiling tube may present as a function of flow rate have been studied. Special features are observed at very low and very high speeds. 3 - Burn-Out: Under steady operating conditions, it is shown that in a uniformly heated channel the burn-out flux as a function of output rate is generally independent of the length. When burn-out is a result of output oscillation, the

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

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    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

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

    International Nuclear Information System (INIS)

    PetrusTakaki, N.

    2012-01-01

    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)

  12. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    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 design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; 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 trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  13. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    International Nuclear Information System (INIS)

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

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

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    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. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  16. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  17. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

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

  18. Electrically Driven Liquid Film Boiling Experiment

    Science.gov (United States)

    Didion, Jeffrey R.

    2016-01-01

    This presentation presents the science background and ground based results that form the basis of the Electrically Driven Liquid Film Boiling Experiment. This is an ISS experiment that is manifested for 2021. Objective: Characterize the effects of gravity on the interaction of electric and flow fields in the presence of phase change specifically pertaining to: a) The effects of microgravity on the electrically generated two-phase flow. b) The effects of microgravity on electrically driven liquid film boiling (includes extreme heat fluxes). Electro-wetting of the boiling section will repel the bubbles away from the heated surface in microgravity environment. Relevance/Impact: Provides phenomenological foundation for the development of electric field based two-phase thermal management systems leveraging EHD, permitting optimization of heat transfer surface area to volume ratios as well as achievement of high heat transfer coefficients thus resulting in system mass and volume savings. EHD replaces buoyancy or flow driven bubble removal from heated surface. Development Approach: Conduct preliminary experiments in low gravity and ground-based facilities to refine technique and obtain preliminary data for model development. ISS environment required to characterize electro-wetting effect on nucleate boiling and CHF in the absence of gravity. Will operate in the FIR - designed for autonomous operation.

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

  20. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

  1. Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Lofaro, R.J.; Aggarwal, S.

    1992-01-01

    The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

  2. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  3. Interim report on the result of the sodium boiling detection benchmark test using BOR-60 reactor noise data

    International Nuclear Information System (INIS)

    Shinohara, Y.; Watanabe, K.; Hayashi, K.

    1989-01-01

    The present paper deals with the second stage of investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe and first results of analysis of data from a series of boiling experiments carried out in the BOR 60 reactor in the USSR. 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. A proposal for in-service boiling monitoring by acoustic means is briefly described. (author). 1 ref., 8 figs, 1 tab

  4. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  5. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  6. Power distribution effects on boiling water reactor stability

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.

    1989-01-01

    The work presented in this paper deals with the effects of spatial power distributions on the stability of boiling water reactors (BWRs). It is shown that a conservative power distribution exists for which the stability is minimal. These results are relevant because they imply that bounding stability calculations are possible and, thus, a worst-possible scenario may be defined for a particular BWR geometry. These bounding calculations may, then, be used to determine the maximum expected limit-cycle peak powers

  7. Self-Sustaining Thorium Boiling Water Reactors

    International Nuclear Information System (INIS)

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-01-01

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  8. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  9. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    International Nuclear Information System (INIS)

    Aksan, S.N.; Stierli, F.; Analytis, G.T.

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

  10. Models and Stability Analysis of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Dorning, John

    2002-01-01

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

  11. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

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

  12. Analysis of CHF enhancement in Subcooled Flow Boiling Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The key factor of CHF improvement is the increase of surface wettability enhancement. Nanoparticles are deposited on the heater surface during the nucleate boiling experiment. H. S. Ahn et al. conducted an internal flow boiling CHF experiment using a micro-structured Zirlo surface. The authors concluded that the flow boiling CHF in the annular flow regime increases with mass flux because of the stability of the liquid film and the liquid replenishment. T. S. Lee et al. conducted the flow boiling CHF experiments using Fe{sub 3}O{sub 4} nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. The effect of the wettability improvement on the CHF can be minimized in relatively low void fraction in slug flow regime. The purpose of our experiment is to investigate the CHF enhancement trend according to exit quality. Existing theoretical CHF model and mechanism were investigated according to the flow regime. CHF experiment in DI water and nanoparticle deposited surface was investigated in mass flux of 1,000 - 5,000 kg/m{sup 2} s and inlet temperature of 40, 60 and 80 .deg. C. To make the similar nanoparticle coating on surface, nanoparticle deposition process was conducted. The experimental results show that CHF enhancement ratio decreased as exit quality decreased and approached to zero.

  13. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

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

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  14. Detection of boiling by Piety's on-line PSD-pattern recognition algorithm applied to neutron noise signals in the SAPHIR reactor

    International Nuclear Information System (INIS)

    Spiekerman, G.

    1988-09-01

    A partial blockage of the cooling channels of a fuel element in a swimming pool reactor could lead to vapour generation and to burn-out. To detect such anomalies, a pattern recognition algorithm based on power spectra density (PSD) proposed by Piety was further developed and implemented on a PDP 11/23 for on-line applications. This algorithm identifies anomalies by measuring the PSD on the process signal and comparing them with a standard baseline previously formed. Up to 8 decision discriminants help to recognize spectral changes due to anomalies. In our application, to detect boiling as quickly as possible with sufficient sensitivity, Piety's algorithm was modified using overlapped Fast-Fourier-Transform-Processing and the averaging of the PSDs over a large sample of preceding instantaneous PSDs. This processing allows high sensitivity in detecting weak disturbances without reducing response time. The algorithm was tested with simulation-of-boiling experiments where nitrogen in a cooling channel of a mock-up of a fuel element was injected. Void fractions higher than 30 % in the channel can be detected. In the case of boiling, it is believed that this limit is lower because collapsing bubbles could give rise to stronger fluctuations. The algorithm was also tested with a boiling experiment where the reactor coolant flow was actually reduced. The results showed that the discriminant D5 of Piety's algorithm based on neutron noise obtained from the existing neutron chambers of the reactor control system could sensitively recognize boiling. The detection time amounts to 7-30 s depending on the strength of the disturbances. Other events, which arise during a normal reactor run like scrams, removal of isotope elements without scramming or control rod movements and which could lead to false alarms, can be distinguished from boiling. 49 refs., 104 figs., 5 tabs

  15. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR...

  16. Fundamental safety-parameter set for boiling water reactors

    International Nuclear Information System (INIS)

    Johnson, C.B.; Mollerus, F.S.; Carmichael, L.A.

    1980-12-01

    A minimum set of parameters is proposed which will indicate the overall safety status of a commercial Boiling Water Reactor. Parameters were selected by identifying those sufficient to determine if functions of fundamental importance to safety are being accomplished. The selected set was subjected to verification by comparison with a broad spectrum of postulated events. Appropriate control room display of the parameter set should assist the operators in determining the safety status of the plant quickly and accurately, even if a plant event is not immediately understood

  17. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    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)

  18. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

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

    1962-08-14

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

  19. Experiments on the fundamental mechanisms of boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Auracher, H. [Technische Universitaet Berlin (Germany). Inst. fuer Energietechnik]. E-mail: auracher@iet.tu-berlin.de; Buchholz, M. [Robert Bosch GmbH (DS/EDS3), Stuttgart (Germany)]. E-mail: Martin.Buchholz@de.bosch.com

    2005-01-15

    The lecture presents a survey of results found by the author and his team during recent years. An experimental technique for precise and systematic measurements of entire boiling curves under steady-state and transient conditions has been developed. Pool boiling experiments for well wetting fluids and fluids with a larger contact angle (FC-72, isopropanol, water) yield single and reproducible boiling curves if the system is clean. However, even minimal deposits on the surface change the heat transfer characteristic and shift the boiling curve with each test run. The situation is different under transient conditions: heating and cooling transients yield different curves even on clean surfaces. Measurements with microsensors give an insight in the two-phase dynamics above the heating surface and the temperature field dynamics above and beneath the surface. Micro thermocouples (38 {mu}m diameter) embedded in the heater (distance to the surface 3.6 {mu}m ), a micro optical probe (tip diameter {approx} 1.5 {mu}m ) and a micro thermocouple probe (tip diameter {approx} 16 {mu}m ), both moveable above the heater surface, are used for these studies. In nucleate boiling, very localized and rapid temperature drops are observed indicating high heat fluxes at the bottom of the bubbles. Already before reaching the critical heat flux (CHF), hot spots occur the size of which increases towards the Leidenfrost point. In the entire transition boiling regime wetting events are observed, but no ones in film boiling. In low heat flux nucleate boiling very small vapor superheats exist in the bubbles and strong superheats in the surrounding liquid. This characteristic change continuously with increasing wall superheat: the liquid surrounding the vapor approaches saturation whereas the vapor becomes more and more superheated. In film boiling the bubbles leaving the vapor film can reach superheats of 30 K or more near the surface (e.g. for isopropanol). The optical probes confirm a liquid

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

    CERN Document Server

    Alam, Tamanna; Jin, Li-Wen

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Fukutomi, S.; Takigawa, Y.

    1992-01-01

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

  2. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    Science.gov (United States)

    Treshow, M.

    1963-04-30

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

  3. Fuel lattice design in boiling water reactors using path relinking

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz, J. J.; Campos, Y.; Perusquia, R.; Montes, J. L.; Hernandez, J. L.

    2006-01-01

    Full text: Full text: A new system for the optimization of fuel lattice design in boiling water reactors (BWR), using the heuristic technique called Path Relinking was developed. The system starts with an initial uranium enrichment and gadolinium percent proposal. With this information, the system generates a seed fuel lattice, which it is used to perform an iterative process until an optimized fuel lattice design is achieved. The iterative process includes two steps. In the first one, we constructed a scatter set with 96 fuel lattices, each fuel lattice we called an element. Starting from this set, we build a reference set with 10 elements, which are the best elements according to the objective function. After, from remaining 86 elements, we build the 10 elements with the maximum distance with respect to reference set. During the iterative process, elements from both sets are used to generate a new element to update the reference set. In the second step, in order to improve the solution achieved up to this moment, two elements from the reference set for constructing new paths beyond to the neighbourhood space, are used. If the new element does not improve the solution, we continue working with the same reference set in the next iteration. The objective function includes both the power peaking factor and the effective multiplication factor at the beginning of the life of the fuel lattice. The principal idea is to minimize the power peaking factor and to keep the effective multiplication factor in a proposal interval. The fuel lattice designed corresponds to the bottom of the fuel assembly. Only, if fuel lattice fulfils the requirements, then it is evaluated at several burnup points. In order to calculate the parameters involved in the objective function the 2D Helios-1.5 code was used. The system was developed in an Alpha Workstation

  4. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

  5. Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2008-01-01

    The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 deg. C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the 'BWRVIP-60 SCC disposition lines' were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the 'BWRVIP-60 SCC disposition lines'

  6. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

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

  7. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY 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...

  8. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... another entity who intends to use the design in some fashion without approval or compensation to the... 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...

  9. Water inventory management in condenser pool of boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.

    1996-01-01

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

  10. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

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

  11. Nonlinear dynamics and chaos in boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.

    1988-01-01

    There are currently 72 commercial boiling water reactors (BWRs) in operation or under construction in the western world, 37 of them in the United States. Consequently, a great effort has been devoted to the study of BWR systems under a wide range of plant operating conditions. This paper represents a contribution to this ongoing effort; its objective is to study the basic dynamic processes in BWR systems, with special emphasis on the physical interpretation of BWR dynamics. The main thrust in this work is the development of phenomenological BWR models suited for analytical studies performed in conjunction with numerical calculations. This approach leads to a deeper understanding of BWR dynamics and facilitates the interpretation of numerical results given by currently available sophisticated BWR codes. 6 refs., 14 figs., 2 tabs

  12. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    International Nuclear Information System (INIS)

    Lanthen, Jonas

    2006-09-01

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes

  13. Mark I 1/5-scale boiling water reactor pressure suppression experiment quick-look report, for test numbers 3.3(a), 3.3(b), 3.4(a), and 3.4(b) performed on May 3, 1977

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    The tests conducted on the 1/5-scale BWR Mark I pressure suppression test facility simulate the three-dimensional transient conditions that are encountered in a wetwell pressure suppression system during a hypothetical loss-of-coolant accident (LOCA). Specifically, the nitrogen (N2)-driven air clearing phase tests discussed here were performed to obtain the air/water-induced dynamic vertical load function and to determine the response of a 90 0 sector of a 360 0 torus structure

  14. Boils

    Science.gov (United States)

    ... as tender, pinkish-red, and swollen, on a firm area of the skin. Over time, it will ... skin areas or joining with other boils Quick growth Weeping, oozing, or crusting Other symptoms may include: ...

  15. A microgravity boiling and convective condensation experiment

    Science.gov (United States)

    Kachnik, Leo; Lee, Doojeong; Best, Frederick; Faget, Nanette

    1987-12-01

    A boiling and condensing test article consisting of two straight tube boilers, one quartz and one stainless steel, and two 1.5 m long glass-in-glass heat exchangers, on 6 mm ID and one 10 mm ID, was flown on the NASA KC-135 0-G aircraft. Using water as the working fluid, the 5 kw boiler produces two phase mixtures of varying quality for mass flow rates between 0.005 and 0.1 kg/sec. The test section is instrumented at eight locations with absolute and differential pressure transducers and thermocouples. A gamma densitometer is used to measure void fraction, and high speed photography records the flow regimes. A three axis accelerometer provides aircraft acceleration data (+ or - 0.01G). Data are collected via an analog-to-digital conversion and data acquisition system. Bubbly, annular, and slug flow regimes were observed in the test section under microgravity conditions. Flow oscillations were observed for some operating conditions and the effect of the 2-G pullout prior to the 0-G period was observed by continuously recording data throughout the parabolas. A total fo 300 parabolas was flown.

  16. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  17. Construction of the advanced boiling water reactor in Japan

    International Nuclear Information System (INIS)

    Natsume, Nobuo; Noda, Hiroshi

    1996-01-01

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

  18. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor 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

  19. Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

    Energy Technology Data Exchange (ETDEWEB)

    Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

    2016-10-15

    Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in

  20. Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

    International Nuclear Information System (INIS)

    Alder, H.P.; Buckley, D.; Grauer, R.; Wiedemann, K.H.

    1992-01-01

    The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. Based on a comprehensive literature study concerning this theme, it has been attempted to identify the individual stages of the activity build-up and to classify their importance. The following areas are discussed in detail: The origins of the corrosion products and of cobalt-59 in the reactor feedwaters; the consolidation of the cobalt in the fuel pins deposits (activation); the release and transport of cobalt-60; the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarized. 90 refs, figs and tabs

  1. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  3. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  4. Multi-cycle boiling water reactor fuel cycle optimization

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  5. Efficient characterization of fuel depletion in boiling water reactor

    International Nuclear Information System (INIS)

    Kim, S.H.

    1980-01-01

    An efficient fuel depletion method for boiling water reactor (BWR) fuel assemblies has been developed for fuel cycle analysis. A computer program HISTORY based on this method was designed to carry out accurate and rapid fuel burnup calculation for the fuel assembly. It has been usefully employed to study the depletion characteristics of the fuel assemblies for the preparation of nodal code input data and the fuel management study. The adequacy and the effectiveness of the assessment of this method used in HISTORY were demonstrated by comparing HISTORY results with more detailed CASMO results. The computing cost of HISTORY typically has been less than one dollar for the fuel assembly-level depletion calculations over the full life of the assembly, in contrast to more than $1000 for CASMO. By combining CASMO and HISTORY, a large number of expensive CASMO calculations can be replaced by inexpensive HISTORY. For the depletion calculations via CASMO/HISTORY, CASMO calculations are required only for the reference conditions and just at the beginning of life for other cases such as changes in void fraction, control rod condition and temperature. The simple and inexpensive HISTORY is sufficienty accurate and fast to be used in conjunction with CASMO for fuel cycle analysis and some BWR design calculations

  6. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.; Srinivasan, V.S.

    1994-01-01

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs

  7. Environmental factors influencing stress corrosion cracking in boiling water reactors

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1984-01-01

    The mechanisms of intergranular stress corrosion cracking (IGSCC) of sensitized stainless steels in boiling water reactor (BWR) primary coolant are reviewed, with emphasis on the role the environment plays on both the initiation and propagation processes. Environmental factors discussed include oxygen (corrosion potential), temperature, and dissolved ions in the water and the range of strain rates at which IGSCC occurs. Both crack propagation rates and the range of strain rates at which IGSCC occurs decrease rapidly as temperature is increased above approximately 200 0 C, in essentially the same manner as the solubility of magnetite decreases in acidic solutions. A mechanism of crack propagation is presented base on this observation. To establish water chemistry guidelines for crack-free operation of BWR's containing sensitized stainless steel, more information is needed on the role of absorption of impurities in the surface and deposited oxides and on the interaction between the oxygen and impurity levels required to maintain an electrochemical potential in a range where IGSCC is unlikely to occur. The relative effects of short bursts of impurities and longer term lower concentrations of these same impurities also need to be evaluated

  8. Status of sodium boiling noise detection programme at reactor research centre, India

    International Nuclear Information System (INIS)

    Prabhakar, R.; Elumalai, G.

    1982-01-01

    Acoustic detection of sodium boiling is a promising technique to monitor subassembly fault in a last reactor. This paper summarises the programme for developing this detection system and describes the design of a high temperature transducer for boiling detection. It is appreciated that the background noise from primary pumps can interfere with this detection. Noise measurements were therefore carried out during water testing of the primary pump of the Fast Breeder Test Reactor. Some preliminary results of these measurements are presented

  9. Calculation of a pressurized-water reactor and a boiling-water reactor fuel rod cluster using the finite element method with first order triangular elements

    International Nuclear Information System (INIS)

    Birkhold, U.; Schmidt, F.A.R.

    1975-07-01

    The FEM-2D programme was used to solve the two-dimensional, time-independent diffusion equation in multi-group form. FEM-2D stands for Finite Element Method two-dimensional Diffusion. Triangular elements with linear flow statement were chosen to describe the given geometrical figure - a pressurized-water reactor (PWR) type Biblis and a boiling-water reactor fuel rod cluster with 5 x 5 fuel rods. Calculations were performed with 301 and 1,204 elements in the pressurized-water reactor, and the boiling-water reactor fuel rod cluster with 900 or 1,296 elements. Calculations with FEM-2D with triangular elements of the 2nd order and calculations of the KWK with the computer programmes MEDIUM and EXTERMINATOR for the PWR or PDQ for the BWR fuel rod cluster were available for comparison. The results were most satisfactory. (orig./LH) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  11. Three-dimensional calculation of boiling water reactors with the FLARE-EIR program

    Energy Technology Data Exchange (ETDEWEB)

    Maeder, C.; Varadi, G.

    1973-06-01

    FLARE-EIR is a modified version of the boiling water reactor simulator FLARE containing a new formula for the nodal coupling coefficient which is based on the one-dimensional one-group response matrix theory. The boundary conditions are represented by albedos that have a clear physical meaning. The quality of the new physics model is investigated by means of three simple test examples which are calculated with diffusion theory in three and one energy groups, with FLARE-EIR and with the original FLARE method. FLARE-EIR contains a new thermohydraulic model for the calculation of the void and mass flow distributions. This model does not require fit parameters that have to be determined with separate computer programs as in the original FLARE version. As a test for FLARE- EIR a startup experiment of the Muehleberg reactor is recalculated, and the results are compared with measurements. (auth)

  12. Development and testing of high-performance fuel pin simulators for boiling experiments in liquid metal flow

    International Nuclear Information System (INIS)

    Casal, V.

    1976-01-01

    There are unknown phenomena, about local and integral boiling events in the core of sodium cooled fast breeder reactors. Therefore at GfK depend out-of-pile boiling experiments have been performed using electrically heated dummies of fuel element bundles. The success of these tests and the amount of information derived from them depend exclusively on the successful simulation of the fuel pins by electrically heated rods as regards the essential physical properties. The report deals with the development and testing of heater rods for sodium boiling experiments in bundles including up to 91 heated pins

  13. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

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

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The 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.

  15. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der.

    1989-01-01

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  16. Boiling water reactors stability analysis - a challenge for the study of complex nonlinear dynamical systems

    International Nuclear Information System (INIS)

    Hennig, D.

    1997-01-01

    In boiling water reactors, there is a region in the operating map for which the reactor exhibits stable or unstable power oscillations. This oscillatory behaviour had to be understood in detail, in order to estimate, in a reliable way, the stability limits. This paper describes the BWR stability analysis methodology used at PSI and presents some recent results. (author) figs., tab., 38 refs

  17. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-03-09

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting on March 18...

  18. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Science.gov (United States)

    2010-02-22

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR) The ACRS Subcommittee on ABWR will hold a meeting on March 2, 2010, at 11545...

  19. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    M. Ishii; S. T. Revankar; T. Downar; Y. Xu, H. J. Yoon; D. Tinkler; U. S. Rohatgi

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed

  20. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

    2003-01-01

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  1. Analysis of counterpart tests performed in boiling water reactor experimental simulators

    Energy Technology Data Exchange (ETDEWEB)

    Bovalini, R.; D' Auria, F.; De Varti, A.; Maugeri, P.; Mazzini, M. (Univ. degli Studi di Pisa, Dept. di Construzioni Meccaniche e Nucleari, Via Diotisalvi 2, 56100 Pisa (IT))

    1992-01-01

    In this paper the main results obtained at the University of Pisa on small-break loss-of-coolant accident counterpart experiments carried out in boiling water reactor (BWR) experimental simulators are summarized. In particular, the results of similar experiments performed in the PIPER-ONE, Full Integral Simulation Test (FIST), and ROSA-III facilities are analyzed. The tests simulate a transient originated by a small break in the recirculation line of a BWR-6 with the high-pressure injection systems unavailable. RELAPS/MOD2 nodalizations have been set up for these facilities and for the reference BWR plant. The calculated results are compared among each other and with the experimental data. Finally, the merits and the limitations of such a program are discussed in view of the evaluation of code scaling capabilities and uncertainty.

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

    International Nuclear Information System (INIS)

    Tuerkcan, E.; Kozma, R.; Verhoef, J.P.; Nabeshima, K.

    1996-01-01

    An important goal of nuclear reactor instrumentation is the continuous monitoring of the state of the reactor and the detection of deviations from the normal behaviour at an early stage. Early detection of anomalies enables one to make the necessary steps in order to prevent further damage of nuclear fuel. In the present paper, an on-line core monitoring system is described by means of which boiling anomaly in nuclear reactor fuel assemblies can be detected. (author). 9 refs, 7 figs

  3. Investigations of decay heat removal by natural convection with boiling in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Kaiser, A.; Peppler, W.; Strake, M.

    1979-03-01

    The safety analysis of a LMFBR indicates the requirement of safely removing the decay heat produced after a reactor shut-down, especially in the case of a failure of all primary circuits. To investigate the conditions under which power in the range of the decay heat can be transfered from a pin bundle to a sodium loop by natural convection, a series of experiments was carried out. Special attention was paid to the behaviour of the natural convection system when boiling occurs, and also to the limits of cooling capability. To apply the experimental results a computer program was made using a simplified model of the emergency cooling system of the SNR 300. With this program several cases of emergency cooling under the boundary conditions of in-tank natural convection were analyzed, assuming a breach of a primary circuit. As an example, the consequences of an increase of the flow resistances in a subassembly were investigated. It was demonstrated that under conditions of steady state boiling there will be only very low vapour qualities. Similar results were obtained from investigations when the sodium temperature at the inlet to the core was elevated, and when the flow resistances in the cold leg of the natural convection loop were increased by a factor of two. Further experiments gave evidence that the cooling of the bundle will substantially be maintained under conditions of low vapour qualities. In summary, it may be stated that even under very pessimistic assumptions concerning the progress of the in-tank natural circulation, the cooling will be maintained reliably, even if boiling occurs for some time. (orig.) [de

  4. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    Energy Technology Data Exchange (ETDEWEB)

    Kruijf, W.J.M. de E-mail: kruijf@iri.tudelft.nl; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C

    2003-04-01

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs.

  5. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C.

    2003-01-01

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs

  6. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Alder, H.P.; Schenker, E.

    1993-01-01

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  7. Improvement of the RELAP5 subcooled boiling model for low pressure conditions

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2000-01-01

    The RELAP5/MOD3.2.2 Gamma code was assessed against low pressure subcooled boiling experiments performed by Zeitoun and Shoukri [1] in a vertical annulus. The predictions of subcooled boiling bubbly flow showed that the present version of the RELAP5 code underestimates the void fraction growth along the tube. To improve the void fraction prediction at low pressure conditions a set of model changes is proposed, which includes modifications of bubbly-slug transition criterion, drift-flux model, interphase heat transfer coefficient and wall evaporation modeling. The improved experiment predictions with the modified RELAP5 code are presented and analysed. (author)

  8. Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

    International Nuclear Information System (INIS)

    Dybeck, P.; Gustafsson, B.

    1990-01-01

    This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

  9. Basic philosophy of the safety design of the Toshiba boiling water reactor

    International Nuclear Information System (INIS)

    Sato, T.

    1992-01-01

    This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

  10. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Classification of decommissioning wastes. Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference boiling water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class A, 97.5%; Class B, 2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

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

  12. Failure analysis of cracked head spray piping from the Dresden Unit 2 Boiling Water Reactor

    International Nuclear Information System (INIS)

    Diercks, D.R.; Dragel, G.M.

    1983-07-01

    Several sections of Type 304 stainless steel head spray piping, 6.25 cm (2.5 in.) in diameter, from the Dresden Unit 2 Boiling Water Reactor were examined to determine the nature and causes of coolant leakages detected during hydrostatic tests. Extensive pitting was observed on the outside surface of the piping, and three cracks, all located at a helical stripe apparently rubbed onto the outer surface of the piping, were also noted. Metallographic examination revealed that the cracking had initiated at the outer surface of the pipe, and showed it to be transgranular and highly branched, characteristic of chloride stress corrosion cracking. The surface pitting also appeared to have been caused by chlorides. A scanning electron microprobe x-ray analysis of the corrosion product in the cracks confirmed the presence of chlorides and also indicated the presence of calcium

  13. Determination of local boiling in light water reactors by correlation of the neutron noise

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    1968-01-01

    The power limit of swimming-pool type reactors depends on the phenomenon of the appearance of burn-out. In order to determine this limit we have attempted to detect the local boiling which usually occurs before the burn out. Local boiling has been simulated by an electrically heated plate placed in the core of the reactor Siloette. The study of local boiling, which is based on the properties of the correlation functions for the neutron noise of detectors placed in the core, shows that a privileged frequency occurs in the power spectrum of the noise. It is intended in the future to determine the influence of various parameters on this characteristic frequency. (author) [fr

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

    Science.gov (United States)

    Sorensen, Kirk; Juhasz, Albert

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Nishizawa, Y.; Kiguchi, T.; Kobayashi, S.; Takumi, K.; Tanaka, H.; Tsutsumi, R.; Yokomi, M.

    1982-01-01

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

  16. Nonlinear dynamics and stability of boiling water reactors: qualitative and quantitative analyses

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1985-01-01

    A phenomenological model has been developed to simulate the qualitative behavior of boiling water reactors (BWRs) in the nonlinear regime under deterministic and stochastic excitations. After the linear stability threshold is crossed, limit cycle oscillations appear due to interactions between two unstable equilibrium points and the phase-space trajectories. This limit cycle becomes unstable when the feedback gain exceeds a certain critical value. Subsequent limit cycle instabilities produce a cascade of period-doubling bifurcations that leads to a periodic pulsed behavior. Under stochastic excitations, BWRs exhibit a single characteristic resonance, at approx.0.5 Hz, in the linear regime. By contrast, this work shows that harmonics of this characteristic frequency appear in the nonlinear regime. Furthermore, this work also demonstrates that amplitudes of the limit cycle oscillations do not depend on the variance of the stochastic excitation and remain bounded at all times. A physical model of nonlinear BWR dynamics has also been developed and employed to calculate the amplitude of limit cycle oscillations and their effects on fuel integrity over a wide range of operating conditions in the Vermont Yankee reactor. These calculations have confirmed that, beyond the threshold for linear stability, the reactor's state variable undergo limit cycle oscillations

  17. Use of adaptive diffusion theory based monitors in optimizing boiling water reactor core designs

    International Nuclear Information System (INIS)

    Congdon, S.P.; Martin, C.L.; Crowther, R.L.

    1988-01-01

    Three-dimensional coarse mesh models are routinely used to predict the performance of boiling water reactors. In the adaptive monitory model, the three-dimensional solutions are permanently adapted to incore probe data. The corrections resulting from the adaptive process lead to reliable predictions of future reactor states. The corrections can also be carried forward to future operating cycles. This can shorten the time required to introduce an validate new design and operating strategy improvements. (orig.) [de

  18. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

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

    OpenAIRE

    Winter, Dominik

    2014-01-01

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

  20. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

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

  1. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is... is developing an onsite independent spent fuel storage installation (ISFSI) and plans to move spent... Report that contains Sensitive Unclassified Non- Safeguards Information and is being withheld from public...

  2. Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Vanderhaegen, M.

    2013-01-01

    In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author) [fr

  3. Physical modeling of the boiling crisis: theory and experiment

    International Nuclear Information System (INIS)

    Nikolayev, Vadim; Beysens, Daniel; Chatain, Denis

    2008-01-01

    Full text of publication follows: In this presentation we describe a physical approach to the boiling crisis called also the critical heat flux (CHF) phenomenon. This approach is based on the hypothesis that the boiling crisis is triggered by spreading of individual vapor bubbles over the heater or equivalently by the growth of individual dry spots under the bubbles. The role of bubble coalescence is assumed to be secondary. The spreading is due to forces acting at the microscopic scale, in the neighborhood of the line of triple contact of liquid, vapor and heater where the local heat fluxes are the strongest. This picture is supposed to be independent on boiling conditions. It is confirmed by the pool boiling experiments carried out at extremely high pressures close to the gas-liquid critical point. Such unusual conditions are chosen to slow down the bubble growth sufficiently to be able to observe the dryout dynamics. In the above experiments it lasted during about a minute. To keep the usual bubble geometry, it is necessary to perform such experiments under reduced gravity. The numerical simulations are carried out for high pressures. They show two regimes of bubble growth. When the heat flux is smaller than a threshold value associated with the CHF, a vapor bubble grows and then leaves the heater by buoyancy. When the heat flux is larger than the CHF, the bubble spreads over the heater without leaving it in agreement with the experimental data. This occurs because the vapor recoil force causes both bubble spreading and strong adhesion to the heater. The CHF variation with system parameters predicted by simulations is briefly discussed. (authors) [fr

  4. The Neutronics Design and Analysis of a 200-MW(electric) Simplified Boiling Water Reactor Core

    International Nuclear Information System (INIS)

    Tinkler, Daniel R.; Downar, Thomas J.

    2003-01-01

    A 200-MW(electric) simplified boiling water reactor (SBWR) was designed and analyzed under sponsorship of the U.S. Department of Energy Nuclear Energy Research Initiative program. The compact size of a 200-MW(electric) reactor makes it attractive for countries with a less well developed engineering infrastructure, as well as for developed countries seeking to tailor generation capacity more closely to the growth of their electricity demand. The 200-MW(electric) core design reported here is based on the 600-MW(electric) General Electric SBWR core, which was first analyzed in the work performed here in order to qualify the computer codes used in the analysis. Cross sections for the 8 x 8 fuel assembly design were generated with the HELIOS lattice physics code, and core simulation was performed with the U.S. Nuclear Regulatory Commission codes RELAP5/PARCS. In order to predict the critical heat flux, the Hench-Gillis correlation was implemented in the RELAP5 code. An equilibrium cycle was designed for the 200-MW(electric) core, which provided a cycle length of more than 2 yr and satisfied the minimum critical power ratio throughout the core life

  5. Stability and dynamic performance of the General Electric Boiling Water Reactor. Licensing topical report

    International Nuclear Information System (INIS)

    1977-01-01

    The analytical methods used to evaluate the stability of GE boiling water reactors is presented. The physical and phenomenological characteristics pertinent to stability analyses are described for the following three configurations evaluated in the BWR design process: total plant, core, and channel hydrodynamic characteristics. Given is a description of the stability criteria and its theoretical basis followed by a description of the analytical methods used in evaluating the BWR. These analyses were derived from and supported by test data from current operating General Electric boiling water reactors. In addition, a parametric evaluation of the BWR is made for the total plant, reactor core, and channel hydrodynamic performance over a wide range of operating conditions. The information presented demonstrates the technical proficiency of the design and substantiates the operational stability characteristics of the integrated Nuclear Steam Supply System

  6. The diagnostics of a nuclear reactor by the analysis of boiling sound

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko; Tanaka, Yoshihisa; Ohsawa, Takaaki; Ohta, Masao

    1980-01-01

    This paper is described on the basic research concerning the method of detecting abnormality by analyzing boiling sound when the heat transfer to coolant became locally abnormal in a pressurized nuclear reactor. In this study, the power spectra of sound were grasped as a sort of pattern, and it was aimed at to diagnose exactly the state in a reactor by analyzing the change with an electronic computer. As the calculating method, the theory of linear distinction function was applied. The subcritical experimental apparatus was used as a simulated reactor core vessel, and boiling sound was received with a hydrophone, amplified, digitalized and processed with a computer. The power spectra of boiling sound were displayed on an oscilloscope, and the digital values were stored in a micro-computer. The method of calculating treatment of the power spectra stored as the data in the microcomputer is explained. The magnitude of the power spectra was large in low frequency region, and decreased as the frequency became higher. The experimental conditions and the results are described. According to the results, considerably good distinction capability was obtained. By utilizing the power spectra in relatively low frequency region, the detection of boiling sound can be made with considerably high accuracy. (Kako, I.)

  7. Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

    International Nuclear Information System (INIS)

    Veer, J.H.C. v.d.; Kema, N.V.

    1982-01-01

    Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

  8. The near boiling reactor: design of a small nuclear reactor for extending the operational envelope of the Victoria Class Submarine

    International Nuclear Information System (INIS)

    Cole, C.; Bonin, H.

    2005-01-01

    A small, inherently safe nuclear reactor that will provide enough power to maintain the hotel load of the Victoria Class Submarine and extend her operational envelope, has been conceptually designed. The final reactor concept, named the Near Boiling (NB) Reactor, employs TRISO fuel particles in Zirconium cladded fuel rods. The reactor is light water moderated and cooled. The core life is specifically designed to coincide with the refit cycle of the Victoria Class Submarine. The reactor employs a simple and reliable control and shut down system that requires little intervention on the part of the submarine's crew. Also, a kinetic model is developed that demonstrates the inherent safety features of the reactor during several accident scenarios. (author)

  9. Internals segmentation at Shoreham 850 MWe boiling water reactor

    International Nuclear Information System (INIS)

    Hirsch, D.J.; Garvey, T.F.

    1993-01-01

    The decommissioning of the Shoreham Nuclear Power Station required the removal and segmentation of the reactor internals. PCI Energy Services, a contractor specialising in remote cutting, machining, and welding performed this work. To make the required cuts, twenty-five different tools were designed, fabricated, assembled, and qualified. The tools included an internals work platform for cutting components in the reactor pressure vessel (RPV), a Wet Cutting Station with a 15' long Plasma Arc Cutting torch delivery mast for segmenting components after removal from the Reactor Pressure Vessel, and a Dry Cutting Station for in-air segmenting of less contaminated components. Other tools included remote-operated torch delivery tools, remote vision systems, filtration systems, and airborne contaminant catchment systems. The work started in mid-June 1992 and was completed in November 9 1992, ahead of schedule and below the ALARA goal. (author)

  10. Stress corrosion cracking of ferritic reactor pressure vessel steels under boiling water reactor conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.

    2001-01-01

    The stress corrosion cracking (SCC) behaviour of low-alloy reactor pressure vessel (RPV) steels in oxygenated high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both, RPV structural integrity and safety, has been a subject of controversial discussions for many years. The SCC crack growth behaviour of different RPV steels under simulated BWR/NWC conditions was therefore characterized by constant load and ripple load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. Modern high-temperature water loops, online crack growth monitoring (DCPD) and fractographical analysis by scanning electron microscopy were used to quantify the cracking response. It is concluded that there is no susceptibility to sustained SCC crack growth at temperatures around 288 C under purely static loading, as long as small-scale-yielding conditions prevail at the crack tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water chemistry conditions (EPRI Action Level 3) and/or for highly stressed specimens, either loaded near to K IJ or with a high degree of plasticity in the remaining ligament. The conservative character of the 'BWR VIP 60 Disposition Lines 1 and 2' for SCC crack growth in low-alloy steels has been confirmed by this study for 288 C and RPV base material. Preliminary results indicate, that these disposition lines may be significantly or slightly exceeded (even in steels with low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 - 250 C) in RPV materials, which show a distinct susceptibility to Dynamic Strain Ageing (DSA). (orig.)

  11. Formation and deposition of platinum nanoparticles under boiling water reactor conditions

    Science.gov (United States)

    Grundler, Pascal V.; Veleva, Lyubomira; Ritter, Stefan

    2017-10-01

    Stress corrosion cracking (SCC) is a well-known degradation mechanism for components of boiling water reactors (BWRs). Therefore the mitigation of SCC is important for ensuring the integrity of the reactor system. Noble metal chemical application (NMCA) has been developed by General Electric to mitigate SCC and reduce the negative side-effects of hydrogen water chemistry used initially for SCC mitigation. NMCA is now widely applied as an online process (OLNC) during power operation. However, the understanding of the parameters that control the formation and deposition of the noble metal (Pt) particles in a BWR was still incomplete. To fill this knowledge gap, systematic studies on the formation and deposition behaviour of Pt particles in simulated and real BWR environment were performed in the framework of a research project at PSI. The present paper summarizes the most important findings. Experiments in a sophisticated high-temperature water loop revealed that the flow conditions, water chemistry, the Pt injection rate, and the pre-conditioning of the stainless steel surfaces have an impact on the Pt deposition behaviour. Slower Pt injection rates and stoichiometric excess of H2 over O2 produce smaller particles, which may increase the efficiency of the OLNC technique in mitigating SCC. Surfaces with a well-developed oxide layer retain more Pt particles. Furthermore, the pre- and post-OLNC exposure times play an important role for the Pt deposition on specimens exposed at the KKL power plant. Redistribution of Pt in the plant takes place, but most of the Pt apparently does not redeposit on the steel surfaces in the reactor system. Comparison of lab and plant results also demonstrated that plant OLNC applications can be simulated reasonably well on the lab scale.

  12. Generic risk insights for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Chung, J.

    1991-05-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs

  13. Proceedings of the International Workshop on Boiling Water Reactor Stability

    International Nuclear Information System (INIS)

    1991-01-01

    With regard to technical understanding of the phenomena, the participants agreed that the causes of instability appear to be well understood, but there are many variables involved, and their correlation with instability conditions is not always certain. Most codes claimed to be capable of predicting oscillations and unstable conditions, based on post-test analyses of data from actual events, but there do not seem to be any blind predictions available which accurately predict an instability event before the actual test results are released. As a result, reactor owners have decided that the best course is to avoid, with sufficient margin, certain regions in the power-flow map where regions of instability are known to exist, rather than try to predict them very accurately. The meeting concluded that the safety significance of BWR instability is rather limited, and current estimates of plant risk do not show it to be a dominant contributor. This is because the installed plant protection systems will shut a reactor down when the oscillations exceed power limits, and any fuel damage which might occur will be localized and containable. However, it was also agreed that an instability event could increase uncertainties in the human error rate, because operators who have never seen an unstable reactor may take actions which are not necessarily the best for the particular situation. In addition, although an instability event may not cause any harm to the public, it may cause some fuel failures, and these are certainly a concern to a reactor owner, for economic and radiation protection reasons. Finally, it was also agreed that BWR instability is certainly considered to be significant by the public, where acceptance of the technology would erode if a plant is perceived to be in an uncontrolled state, regardless of the actual risk inherent in the situation

  14. Effect of water impurities on stress corrosion cracking in a boiling water reactor

    International Nuclear Information System (INIS)

    Ljungbery, L.G.; Cubicciotti, D

    1985-01-01

    A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

  15. Studies on improvements in the control methods of boiling water reactor plant

    International Nuclear Information System (INIS)

    Mankin, Shuichi

    1982-08-01

    In order to improve the performance of regulation and load following control of boiling water reactor plant, optimal control theory is applied and new types of control method are developed. Case-α controller is first formulated on the basis of the optimal linear regulator theory applied to the linealized model of the system; it is then modified by adding a integration-type action in a feed back loop and by the use of variable gain and reference for adapting to the power level requested. Case-#betta# controller consists of a hierarchical control scheme which has classical P.I. type sub-loop controllers at the first level and a linear optimal regulator at the second level. The controller is designed on the basis of the optimal regulator theory applied to the multivariate autoregressive system model which is obtained from the identification experiments, where the system model is determined with the conventional sub-loop controllers included. The results of the simulation experiments show these control methods proposed have performed fairly well and will be useful for the improvement of the performance of nuclear power plant control. In addition, it is suggested that these control methods will be also attractive for the control of other production plants because these were developed in the attempt to solve the problems deviated from so called 'The gap between the optimal contro theory and actual systems.' (author)

  16. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  17. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  18. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1995-01-01

    The possible allowance of reactivity credit for the exposure history, of power reactor fuel has spurred interest because of the potential of greatly reduced risk and cost when applied to the design and certification of spent-fuel casks used for transportation and storage. Previous pressurized water reactor feasibility assessments are extended to boiling water reactor fuel

  19. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  20. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  1. Theoretical aspects of the rehocence method application in boiling water reactors

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1982-01-01

    Theoretical aspects of a new correlation method, the so-called ''rehocence'' of ''smoothed coherence transform'', for transit time estimation in boiling water reactors are given in this paper. The used rehocence method presents the transit time directly in the same way in the ordinary cross correlation technique, but with a better resolution, even when the measured signals are contaminated by a narrow-band limited internal noise coming from the global noise of the neutron flux fluctuation. (author)

  2. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    Lee, Bom Soon.

    1994-01-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  3. Investigation of boiling water reactor stability and limit-cycle amplitude

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.A.; Euler, J.A.

    1991-01-01

    Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

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

    International Nuclear Information System (INIS)

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

    1989-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

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

  6. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

    1995-01-01

    Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

  7. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  8. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-04-30

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

  10. Preliminary Study on CHF Enhancement of Cellulose Nano Fiber (CNF) Fluid with Wire Pool Boiling Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Won Ki; Lee, Yun Seok; Lim, Dong Young; Song, Sub Lee; Lee, Jae Young; Lee, Kwon Yeong [Hanyang Global University, Pohang (Korea, Republic of); Hwang, Dong Soo [POSTECH, Pohang (Korea, Republic of)

    2016-05-15

    Critical heat flux (CHF) is enhancement of a boiling system will make more compact and effective cooling systems, for examples, nuclear reactors, and air conditioning units. For decades, researchers have been trying to develop more efficient working fluid for heat transfer. This is where nano-fluid could play a key role. There have been a lot of researches for CHF enhancements in nucleate boiling by using nano-fluid which are composed of metal such as copper, Al{sub 2}O{sub 3} and ceramic. And a critical factor of the enhancement is deposition of nano-particles on heating surface, although some results of recent studies are contrary. Also, previous nano-fluid are expensive and have a problem in mass production, so they are difficult to apply to practical industries. Therefore we chose a new material, cellulose nano fiber (CNF) as a solution. CNF can be applied to real situation because it has some advantages which are cost-effectiveness, easiness to get and to make it in nano scale. CHF performance of CNF fluid was different from that of distilled water. Compared to CHF of distilled water, CHF of the CNF fluid which had 0.001V%, 0.01V%, and 0.1V% volumetric concentrations were enhanced to 1%, 104%, and 13% respectively. Likewise other nano-fluid, deposition phenomena was observed in this CNF fluid boiling experiment.

  11. Phenomenology and thermo-hydraulic stability of the CAREM-25 reactor: Evaluation of subcooled boiling effect

    International Nuclear Information System (INIS)

    Acuna, F.M.; Marcel, C.P.; Zanocco, P.G.; Delmastro, D.F.

    2012-01-01

    In this article the phenomenology present in self/pressurized, integral, natural circulation, low thermodynamic quality nuclear reactors similar to CAREM-25 is investigated. In particular, analytical relations for the mass flow rate, the core mean enthalpy and the location of the two phase boundary are derived in terms of the so-called natural variables of the system: the nuclear power, the condensation power and the system pressure. In addition, some consequences of the flashing phenomenon in the reactor thermal-hydraulics are discussed emphasizing those affecting the reactor stability. The reactor stability performance was studied by using the HUARPE code which is a low diffusive code. The stability results obtained by neglecting the subcooled effect in the system are presented in the so-called the stability maps in which the results are presented for a wide range of conditions. The stability effect caused by the presence of subcooled boiling in the reactor core was also examined. In order to investigate such a consequence, the code was slightly modified such that the predicted vapor profile in the hot leg is similar to that estimated by RELAP system code at steady state conditions. The simple implemented algorithm allows varying a free parameter with which a broad number of cases can be studied. This is important since the subcooled boiling predictions generally have large uncertainties and therefore to cover a large number of situations is desired to derive confident conclusions. The results show the existence of vapor created by means of subcooled boiling enhances the system stability for a wide range of conditions. For this reason from this preliminary investigation, it is concluded neglecting the subcooled effect in CAREM-25 stability studies is a conservative criterion (author))

  12. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

  13. Remote mechanized equipment for the repair and replacement of boiling water reactor recirculation loop piping

    International Nuclear Information System (INIS)

    Mauser, D.; Busch, D.F.

    1983-01-01

    Equipment has been assembled for the remote repair or replacement of boiling water reactor nuclear plant piping in the diameter range of 4 to 28 inches (10-71 cm). The objectives of this program were to produce high-quality pipe welds, reduce plant downtime, and reduce man-rem exposure. The repair strategy was to permit repair personnel to install and check out the repair subsystems and then leave the radiation zone allowing the operations to be conducted at a distance of up to 300 feet (91 m) from the operator. The complete repair system comprises subsystems for pipe severing, dimensional gaging, joint preparation, counterboring, welding, postweld nondestructive inspection (conceptual design), and audio, electronic, and visual monitoring of all operations. Components for all subsystems, excluding those for postweld nondestructive inspection, were purchased and modified as needed for integration into the repair system. Subsystems were designed for two sizes of Type 304 stainless steelpipe. For smaller, 12-inch-diameter (30.5 cm) pipe, severing is accomplished by a power hack saw and joint preparation and counterboring by an internally mounted lathe. The 22-inch-diameter (56 cm) pipe is severed, prepared, and counterbored using an externally mounted, single-point machining device. Dimensional gaging is performed to characterize the pipe geometry relative to a fixed external reference surface, allowing the placement of the joint preparation and the counterbore to be optimized. For both pipe sizes, a track-mounted gas tungsten-arc welding head with filler wire feed is used

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

    Science.gov (United States)

    Nahra, Henry; Hasan, Mohammad; Balasubramaniam, R.; Patania, Michelle; Hall, Nancy; Wagner, James; Mackey, Jeffrey; Frankenfield, Bruce; Hauser, Daniel; Harpster, George; hide

    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.

  15. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik [Nuclear Physics and Biophysics Research Division Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

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

  16. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  17. Identification of neutron noise sources in a boiling water reactor

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

    1977-01-01

    Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

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

    International Nuclear Information System (INIS)

    Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

    2015-01-01

    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/m 2 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 [ru

  19. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  20. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  1. Electrochemical measurements and modeling predictions in boiling water reactors under various operating conditions

    International Nuclear Information System (INIS)

    Indig, M.E.

    1991-01-01

    One important issue for providing life extension to operating boiling water nuclear reactors (BWRs) is the control of stress corrosion cracking in all sections of the primary coolant circuit. This paper links experimental and theoretical methods that provide understanding and measurements of the critical parameter, the electrochemical potential (ECP), and its application to determining crack growth rate among and within the family of BWRs. Measurement of in-core ECP required the development of a new family of radiation-resistant sensors. With these sensors, ECPs were measured in the core and piping of two operating BWRs. Concurrent crack growth measurements were used to benchmark a crack growth prediction algorithm with measured ECPs

  2. Calculation of BWR [Boiling Water Reactor] limit cycle amplitude using Galerkin's method

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.; Euler, J.A.

    1990-01-01

    This paper describes the application of Galerkin's method to estimate the amplitude of boiling water reactor (BWR) limit cycle oscillations. It will be shown that Galerkin's method can be applied to a model of BWR dynamics consisting of the point kinetics equations and the LAPUR generated feedback transfer function to calculate the time history of small amplitude limit cycles. This allows results from the linear frequency domain code LAPUR to be used to calculate nonlinear time domain information. 2 refs., 2 figs., 1 tab

  3. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  4. Development of Flow Boiling and Condensation Experiment on the International Space Station- Normal and Low Gravity Flow Boiling Experiment Development and Test Results

    Science.gov (United States)

    Nahra, Henry K.; Hall, Nancy R.; Hasan, Mohammad M.; Wagner, James D.; May, Rochelle L.; Mackey, Jeffrey R.; Kolacz, John S.; Butcher, Robert L.; Frankenfield, Bruce J.; Mudawar, Issam; hide

    2013-01-01

    Flow boiling and condensation have been identified as two key mechanisms for heat transport that are vital for achieving weight and volume reduction as well as performance enhancement in future space systems. Since inertia driven flows are demanding on power usage, lower flows are desirable. However, in microgravity, lower flows are dominated by forces other than inertia (like the capillary force). It is of paramount interest to investigate limits of low flows beyond which the flow is inertial enough to be gravity independent. One of the objectives of the Flow Boiling and Condensation Flight Experiment sets to investigate these limits for flow boiling and condensation. A two-phase flow loop consisting of a Flow Boiling Module and two Condensation Modules has been developed to experimentally study flow boiling condensation heat transfer in the reduced gravity environment provided by the reduced gravity platform. This effort supports the development of a flow boiling and condensation facility for the International Space Station (ISS). The closed loop test facility is designed to deliver the test fluid, FC-72 to the inlet of any one of the test modules at specified thermodynamic and flow conditions. The zero-g-aircraft tests will provide subcooled and saturated flow boiling critical heat flux and flow condensation heat transfer data over wide range of flow velocities. Additionally, these tests will verify the performance of all gravity sensitive components, such as evaporator, condenser and accumulator associated with the two-phase flow loop. We will present in this paper the breadboard development and testing results which consist of detailed performance evaluation of the heater and condenser combination in reduced and normal gravity. We will also present the design of the reduced gravity aircraft rack and the results of the ground flow boiling heat transfer testing performed with the Flow Boiling Module that is designed to investigate flow boiling heat transfer and

  5. Application of reliability techniques to prioritize BWR [boiling water reactor] recirculation loop welds for in-service inspection

    International Nuclear Information System (INIS)

    Holman, G.S.

    1989-12-01

    In January 1988 the US Nuclear Regulatory Commission issued Generic Letter 88-01 together with NUREG-0313, Revision 2, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' to implement NRC long-range plans for addressing the problem of stress corrosion cracking in boiling water reactor piping. NUREG-0313 presents guidelines for categorizing BWR pipe welds according to their SCC condition (e.g., presence of known cracks, implementation of measures for mitigating SCC) as well as recommended inspection schedules (e.g., percentage of welds inspected, inspection frequency) for each weld category. NUREG-0313 does not, however, specify individual welds to be inspected. To address this issue, the Lawrence Livermore National Laboratory developed two recommended inspection samples for welds in a typical BWR recirculation loop. Using a probabilistic fracture mechanics model, LLNL prioritized loop welds on the basis of estimated leak probabilities. The results of this evaluation indicate that riser welds and bypass welds should be given priority attention over other welds. Larger-diameter welds as a group can be considered of secondary importance compared to riser and bypass welds. A ''blind'' comparison between the probability-based inspection samples and data from actual field inspections indicated that the probabilistic analysis generally captured the welds which the field inspections identified as warranting repair or replacement. Discrepancies between the field data and the analytic results can likely be attributed to simplifying assumptions made in the analysis. The overall agreement between analysis and field experience suggests that reliability techniques -- when combined with historical experience -- represent a sound technical basis on which to define meaningful weld inspection programs. 13 refs., 8 figs., 5 tabs

  6. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-02

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

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

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

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

  9. Final environmental statement for La Crosse Boiling Water Reactor: (Docket No. 50-409)

    International Nuclear Information System (INIS)

    1980-04-01

    A Final Environmental Statement for the Dairyland Power Cooperative for the conversion from a provisional to a full-term operating license for the La Crosse Boiling Water Reactor, located in Vernon County, Wisconsin, has been prepared by the Office of Nuclear Reactor Regulation. This statement provides a summary of environmental impacts and adverse effects of operation of the facility, and a consideration of principal alternatives (including removal of LACBWR from service, alternative cooling methodology, and alternative waste treatment systems). Also included are the comments of federal, state, and local governmental agencies and certain non-governmental organizations on the La Crosse Draft Environmental Statement and staff responses to these comments. After weighing environmental, economic, and technical benefits and liabilities, the staff recommends conversion from a provisional operating license to a full-term operating license, subject to specific environmental protection limitations. An operational monitoring program shall be established as part of the Environmental Technical Specifications. 64 refs., 20 figs., 48 tabs

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

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

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

  11. Passive containment cooling system performance in the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

    1997-01-01

    The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

  12. Conceptual design considerations for a boiling water reactor of the next century

    International Nuclear Information System (INIS)

    Fennern, L.E.; Dillmann, C.W.; Moriya, K.; Murase, M.; Tanabe, A.; Saito, T.; Matsumura, K.; Horimizu, A.

    1993-01-01

    The Advanced Boiling Water Reactor (ABWR) plant has been developed by General Electric (GE), Hitachi and Toshiba under the sponsorship of Tokyo Electric Power Company (TEPCO) and other Japanese utilities. It is currently under construction as the sixth and seventh units at Kashiwazaki Kariwa Power Station. While the ABWR represents a major improvement in operability and economy over designs currently in operation, the future BWR for the next century is desired to be further developed and improved with emphasis on easier operation, maintenance and flexibility of el cycle in order to meet the future social environment. GE, Hitachi, Toshiba and the Japanese utilities, therefore, have begun to develop the concept of the next century BWR. This paper describes the concepts under consideration of the reactor for the future BWR

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

    International Nuclear Information System (INIS)

    Koncar, B.; Mramor, K.

    2007-01-01

    A three-dimensional two-fluid code NEPTUNE C FD 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 NEPTUNE C FD 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)

  14. Application of Galerkin's method for calculating boiling water reactor limit-cycle amplitude using the LAPUR feedback-transfer function and the point-kinetics equations

    International Nuclear Information System (INIS)

    Damiano, B.; March-Leuba, J.A.; Euler, J.A.

    1990-01-01

    This paper describes a technique for calculating boiling water reactor (BWR) behavior during steady-state limit-cycle oscillations. An approximate solution is obtained from the application of Galerkin's method to a BWR dynamic model consisting of the point-kinetics equations and the LAPUR-calculated power-to-reactivity feedback-transfer function. The approximate-solution technique is described, and comparisons of approximate solutions with numerical results and measured data are given. 7 refs., 5 figs

  15. A nuclear desalination complex with a VK-300 boiling type reactor facility

    International Nuclear Information System (INIS)

    Kuznetzov, Y.N.; Mishanina, Y.A.; Romenkov, A.A.

    2004-01-01

    RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m 3 /day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

  16. Application of the Isotope Ratio Method to a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Meriwether, George H.; Mitchell, Mark R.; Reid, Bruce D.

    2010-01-01

    production in a boiling water reactor fuel bundle based on measurements taken from the corresponding fuel assembly channel. Our preliminary results are in good agreement with the actual operating history of the reactor during the cycle that the fuel bundle was resident in the core.

  17. TARMS, an on-line boiling water reactor operation management system

    International Nuclear Information System (INIS)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-01-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

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

  19. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  20. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-08

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

  2. Branch-and-Bound algorithm applied to uncertainty quantification of a Boiling Water Reactor Station Blackout

    International Nuclear Information System (INIS)

    Nielsen, Joseph; Tokuhiro, Akira; Hiromoto, Robert; Tu, Lei

    2015-01-01

    Highlights: • Dynamic Event Tree solutions have been optimized using the Branch-and-Bound algorithm. • A 60% efficiency in optimization has been achieved. • Modeling uncertainty within a risk-informed framework is evaluated. - Abstract: Evaluation of the impacts of uncertainty and sensitivity in modeling presents a significant set of challenges in particular to high fidelity modeling. Computational costs and validation of models creates a need for cost effective decision making with regards to experiment design. Experiments designed to validate computation models can be used to reduce uncertainty in the physical model. In some cases, large uncertainty in a particular aspect of the model may or may not have a large impact on the final results. For example, modeling of a relief valve may result in large uncertainty, however, the actual effects on final peak clad temperature in a reactor transient may be small and the large uncertainty with respect to valve modeling may be considered acceptable. Additionally, the ability to determine the adequacy of a model and the validation supporting it should be considered within a risk informed framework. Low fidelity modeling with large uncertainty may be considered adequate if the uncertainty is considered acceptable with respect to risk. In other words, models that are used to evaluate the probability of failure should be evaluated more rigorously with the intent of increasing safety margin. Probabilistic risk assessment (PRA) techniques have traditionally been used to identify accident conditions and transients. Traditional classical event tree methods utilize analysts’ knowledge and experience to identify the important timing of events in coordination with thermal-hydraulic modeling. These methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical

  3. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

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

  5. Water chemistry in boiling water reactors - A Leibstadt-specific overview

    International Nuclear Information System (INIS)

    Sarott, F.-A.

    2005-01-01

    The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

  6. Investigation and examination on the cracking of pipings in boiling water reactors

    International Nuclear Information System (INIS)

    1977-01-01

    This is the report made by the Reactor Safety Technology Expert Committee to the Atomic Energy Commission regarding the investigation and examination on stress corrosion cracking which seems to be the cause of the cracking of pipings in boiling water reactors, the measures to reduce it, and the subjects of research hereafter. Recently, the stress corrosion cracking of primary coolant pipings has been often observed, and this phenomenon occurred in the pressure boundary of primary coolant, consequently it is possible to be linked to the troubles of large scale. The Reactor Material Subcommittee was established on May 14, 1975, and investigated the cracking phenomena in the recirculating system and core spray system of BWRs in Japan and foreign countries. The recent cases have been concentrated to the heat-affected part due to welding of 304 type austenitic stainless steel pipings of from 4 in to 10 in diameter for BWRs. They are the stress corrosion cracking at grain boundaries occurred under the loaded condition and in the environment of high temperature, high pressure water. The cracking of this kind was never experienced in PWRs. The results of the technical examination, the consideration of the mechanism of stress corrosion cracking, and the countermeasures are described. (Kako, I.)

  7. Approximation model of three-dimensional power distribution in boiling water reactor using neural networks

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2001-01-01

    Fast and accurate prediction of three-dimensional (3D) power distribution is essential in a boiling water reactor (BWR). The prediction method of 3D power distribution in BWR is developed using the neural network. Application of the neural network starts with selecting the learning algorithm. In the proposed method, we use the learning algorithms based on a class of Quasi-Newton optimization techniques called Self-Scaling Variable Metric (SSVM) methods. Prediction studies were done for a core of actual BWR plant with octant symmetry. Compared to classical Quasi-Newton methods, it is shown that the SSVM method reduces the number of iterations in the learning mode. The results of prediction demonstrate that the neural network can predict 3D power distribution of BWR reasonably well. The proposed method will be very useful for BWR loading pattern optimization problems where 3D power distribution for a huge number of loading patterns (LPs) must be performed. (author)

  8. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    International Nuclear Information System (INIS)

    Schlereth, J.R.; Pennington, D.

    1996-01-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it's Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components

  9. Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

    International Nuclear Information System (INIS)

    Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R.

    1996-01-01

    Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder

  10. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  11. Boiling water reactor containment modeling and analysis at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Holcomb, E.E. III; Wilson, G.E.

    1984-01-01

    Under the auspices of the United States Nuclear Regulatory Commission, severe accidents are being studied at the Idaho National Engineering Laboratory. The boiling water reactor (BWR) studies have focused on postulated anticipated transients without scram (ATWS) accidents which might contribute to severe core damage or containment failure. A summary of the containment studies is presented in the context of the analytical tools (codes) used, typical transient simulation results and the need for prototypical containment data. All of these are related to current and future analytical capabilities. It is shown that torus temperatures during the ATWS depart from limiting conditions for BWR T-quencher operation, outside of which stable steam condensation has not been proven

  12. A study of implementing In-Cycle-Shuffle strategy to a decommissioning boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chung-Yuan, E-mail: tuckjason@iner.gov.tw; Tung, Wu-Hsiung; Yaur, Shyun-Jung

    2017-06-15

    Highlights: • A loading pattern strategy ICS (In-Cycle-Shuffle) was implemented to the last cycle of the boiling water reactor. • The best power sharing distribution and ICS timing was found. • A new parameter “Burnup sharing” is presented to evaluate ICS strategy. - Abstract: In this paper, a loading pattern strategy In-Cycle-Shuffle (ICS) is implemented to the last cycle of the boiling water reactor (BWR) before decommissioning to save the fuel cycle cost. This method needs a core shutdown during the operation of a cycle to change the loading pattern to gain more reactivity. The reactivity model is used to model the ICS strategy in order to find out the best ICS timing and the optimum power sharing distribution before ICS and after ICS. Several parameters of reactivity model are modified and the effect of burnable poison, gadolinium (Gd), is considered in this research. Three cases are presented and it is found that the best ICS timing is at about two-thirds of total cycle length no matter the poisoning effect of Gd is considered or not. According to the optimum power sharing distribution result, it is suggested to decrease the once burnt power and increase the thrice burnt fuel power as much as possible before ICS. After ICS, it is suggested to increase the positive reactivity fuel power and decrease the thrice burnt fuel power as much as possible. A new parameter “Burnup sharing” is presented to evaluate the special case whose EOC power weighting factor and the burnup accumulation factor in the reactivity model are quite different.

  13. Analysis of pressure oscillations and safety relief valve vibrations in the main steam system of a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Galbally, David, E-mail: dgalbally@innomerics.com [Innomerics, Calle San Juan de la Cruz 2, 28223 Madrid (Spain); García, Gonzalo [Alava Ingenieros, Calle Albasanz 16, 28037 Madrid (Spain); Hernando, Jesús; Sánchez, Juan de Dios [Iberdrola, Calle Tomás Redondo 1, 28033 Madrid (Spain); Barral, Marcos [Alava Ingenieros, Calle Albasanz 16, 28037 Madrid (Spain)

    2015-11-15

    Highlights: • We analyze the vibratory response of safety relief valves in the main steam system of a Boiling Water Reactor. • We show that valve internals experience acceleration spikes of more than 20 g. • Spikes are caused by impacts between the valve disc and the seating surface of the valve nozzle. • Resonances occur at higher Strouhal numbers than those reported in the literature for tandem side branches. • Valves experience high vibration levels even for resonances caused by second order hydrodynamic modes. - Abstract: Steam flow inside the main steam lines of a Boiling Water Reactor can generate high-amplitude pressure oscillations due to coupling between the separated shear layer at the mouth of the safety relief valves (SRVs) and the acoustic modes of the side branches where the SRVs are mounted. It is known that certain combinations of flow velocities and main steam line geometries are capable of generating self-excited pressure oscillations with very high amplitudes, which can endanger the structural integrity of main steam system components, such as safety valves, or reactor internals such as steam dryers. However, main steam systems may also experience lower amplitude pressure oscillations due, for example, to coupling of higher order hydrodynamic modes with acoustic cavity modes, or to incipient resonances where the free stream velocity is slightly lower than the critical flow velocity required to develop a stable locked-on acoustic resonance. The amplitude of these pressure oscillations is typically insufficient to cause readily observable structural damage to main steam system components, but may still have subtle effects on safety relief valves. The investigation presented in this article focuses on the characterization of the response of SRVs under the effects of pressure oscillations associated with acoustic excitations that are insufficient to cause structural damage to the valves or associated equipment. It is shown that valve

  14. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Hartmann C.

    2018-01-01

    Full Text Available Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc. during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE. Two sets of floating linear voltage differential transformer (LVDT pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has

  15. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

  16. Heat transfer study of a submerged reactor channel under boil-off condition

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

    2012-12-15

    Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

  17. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  18. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-01-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

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

    Energy Technology Data Exchange (ETDEWEB)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

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

    International Nuclear Information System (INIS)

    1994-01-01

    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

  1. Neutron-Detection Based Monitoring of Void Effects in Boiling Water Reactors

    Science.gov (United States)

    Loberg, John; Österlund, Michael; Bejmer, Klaes-Håkan; Blomgren, Jan

    2009-08-01

    The ratio between the thermal and fast neutron flux in a BWR (Boiling Water Reactor) depends on the void fraction. The density of the steam-water mixture present in the core determines the efficiency of the moderation of fast neutrons born in fission, and therefore the void fraction could be determined by means of a simultaneous measurement of the thermal and fast neutron fluxes. Such measurement could also be used to investigate channel bow of the nuclear fuel bundles surrounding the detector because of sensitivity of the thermal flux to geometry changes. Calculations have been performed with lattice codes to study the behavior of the void fraction correlation to the ratio of the thermal and fast neutron flux. The results prove the correlation to be nearly linear and robust. The rate of change of the correlation is insensitive to standard reactor operating parameters such as control rods and burnable absorbers; the sensitivity of the ratio to void fraction changes primarily depends on the geometry of the fuel bundles.

  2. Feasibility of core management system by data communication for boiling water reactors

    International Nuclear Information System (INIS)

    Motoda, H.; Tanisaka, S.; Kiguchi, T.; Yonenaga, H.

    1977-01-01

    A core management system by data communication has been designed and proposed for more efficient operation of boiling water reactor (BWR) plants by faster transmission and centralized management of information. The system comprises three kinds f computers: a process computer for monitoring purposes at the reactor site, a center computer for administration purposes at the head office, and a large scientific computer for planning and evaluation purposes. The process and the large computers are connected to the center computer by a data transmission line. To demonstrate the feasibility of such a system, the operating history evaluation system, which is one of the subsystems of the core management system, has been developed along the above concept. Application to the evaluation of the operating history of a commercial BWR shows a great deal of merit. Quick response and a significant manpower reduction can be expected by data communication and minimized intervention of human labor. Visual display is also found to be very useful in understanding the core characteristics

  3. Bayesian optimization analysis of containment-venting operation in a boiling water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Xiaoyu; Ishikawa, Jun; Sugiyama, Tomoyuki; Maryyama, Yu [Nuclear Safety Research Center, Japan Atomic Energy Agency, Ibaraki (Japan)

    2017-03-15

    Containment venting is one of several essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach to venting operations, from a simulation-based perspective, using an integrated severe accident code, THALES2/KICHE. The effectiveness of the containment-venting strategies needs to be verified via numerical simulations based on various settings of the venting conditions. The number of iterations, however, needs to be controlled to avoid cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using a Gaussian process regression, a surrogate model of the “black-box” code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. Compared with the case of pure random searches, the number of code queries is largely reduced for the optimum finding. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

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

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

    International Nuclear Information System (INIS)

    1993-01-01

    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

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

    Science.gov (United States)

    Cole, Christopher J. P.

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

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

    International Nuclear Information System (INIS)

    Cole, C.J.P.

    2005-01-01

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

  10. Topics to be covered in safety analysis reports for nuclear power plants with pressurized water reactors or boiling water reactors in the F.R.G

    International Nuclear Information System (INIS)

    Kohler, H.A.G.

    1977-01-01

    This manual aims at defining the standards to be used in Safety Analysis Reports for Nuclear Power Plants with Pressurized Water Reactors or Boiling Water Reactors in the Federal Republic of Germany. The topics to be covered are: Information about the site (geographic situation, settlement, industrial and military facilities, transport and communications, meteorological conditions, geological, hydrological and seismic conditions, radiological background), description of the power plant (building structures, safety vessel, reactor core, cooling system, ventilation systems, steam power plant, electrical facilities, systems for measurement and control), indication of operation (commissioning, operation, safety measures, radiation monitoring, organization), incident analysis (reactivity incidents, loss-of-coolant incidents, external impacts). (HP) [de

  11. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

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

    International Nuclear Information System (INIS)

    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 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

  13. Dynamic behavior of large oxide-fueled fast reactors during over-power transients due to boiling of sodium

    International Nuclear Information System (INIS)

    Pires, L.F.G.

    1983-01-01

    The dynamic behavior of large oxide-fueled fast reactors during over-power transients or under-flow situations which result in boiling of sodium used as coolant. The fuel heat transfer was analysed to determine the fuel temperature profile and the Doppler feedback reactivity. The sodium pressure, temperature, mass flow rate and sodium voiding reactivity were obtained by solving the basic coolant hydrodynamics equations. (Author) [pt

  14. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  15. Study of spray cooling of a pressure vessel head of a boiling water reactor

    International Nuclear Information System (INIS)

    Anglart, Henryk; Alavyoon, Farid; Novarini, Remi

    2010-01-01

    The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.

  16. Optimization of fuel exchange machine operation for boiling water reactors using an artificial intelligence technique

    International Nuclear Information System (INIS)

    Sekimizu, K.; Araki, T.; Tatemichi, S.I.

    1987-01-01

    Optimization of fuel assembly exchange machine movements during periodic refueling outage is discussed. The fuel assembly movements during a fuel shuffling were examined, and it was found that the fuel assembly movements consist of two different movement sequences;one is the ''PATH,'' which begins at a discharged fuel assembly and terminates at a fresh fuel assembly, and the other is the ''LOOP,'' where fuel assemblies circulate in the core. It is also shown that fuel-loading patterns during the fuel shuffling can be expressed by the state of each PATH, which is the number of elements already accomplished in the PATH actions. Based on this fact, a scheme to determine a fuel assembly movement sequence within the constraint was formulated using the artificial intelligence language PROLOG. An additional merit to the scheme is that it can simultaneously evaluate fuel assembly movement, due to the control rods and local power range monitor exchange, in addition to normal fuel shuffling. Fuel assembly movements, for fuel shuffling in a 540-MW(electric) boiling water reactor power plant, were calculated by this scheme. It is also shown that the true optimization to minimize the fuel exchange machine movements would be costly to obtain due to the number of alternatives that would need to be evaluated. However, a method to obtain a quasi-optimum solution is suggested

  17. A diagnostic expert system for a boiling water reactor using a dynamic model

    International Nuclear Information System (INIS)

    Sonoda, Y.; Kanemoto, S.; Imaruoka, H.

    1990-01-01

    A diagnostic expert system for abnormal disturbances in a BWR (Boiling Water Reactor) plant has been developed. The peculiar feature of this system is a diagnostic method which combines artificial intelligence technique with numerical analysis technique. The system has three diagnostic functions, 1) identification of anomaly position (device or sensor), 2) identification of anomaly mode and 3) identification of anomaly cause. Function 1) is implemented as follows. First, a hypothesis about anomaly propagation paths is built up by qualitative reasoning, using knowledge of causal relations among observed signals. Next, the abnormal device or sensor is found by applying model reference method and fuzzy set theory to test the hypothesis, using knowledge of plant structure and function, heuristic strategy of diagnosis and module type dynamic simulator. This simulator is composed of basic transfer function modules. The simulation model for the testing region is built up automatically, according to the requirement from the diagnostic task. Function 2) means identification of dynamic characteristics for an anomaly. It is realized by tuning model parameters so as to reproduce the abnormal signal behavior using the non-linear programing method. Function 3) derives probable anomaly causes from heuristic rules between anomaly mode and cause. A basic plant dynamic model was built up and adjusted to dynamic characteristics for one BWR plant (1100MWe). In order to verify the diagnostic functions of this system, data for several abnormal events was compiled by modifying this model. The diagnostic functions were proved useful, through the simulated abnormal data

  18. Higher order generalized perturbation theory for boiling water reactor in-core fuel management optimization

    International Nuclear Information System (INIS)

    Moore, B.R.; Turinsky, P.J.

    1998-01-01

    Boiling water reactor (BWR) loading pattern assessment requires solving the two-group, nodal form of the neutron diffusion equation and drift-flux form of the fluid equations simultaneously because these equation sets are strongly coupled via nonlinear feedback. To reduce the computational burden associated with the calculation of the core attributes (that is, core eigenvalue and thermal margins) of a perturbed BWR loading pattern, the analytical and numerical aspects of a higher order generalized perturbation theory (GPT) method, which correctly addresses the strong nonlinear feedbacks of two-phase flow, have been established. Inclusion of Jacobian information in the definition of the generalized flux adjoints provides for a rapidly convergent iterative method for solution of the power distribution and eigenvalue of a loading pattern perturbed from a reference state. Results show that the computational speedup of GPT compared with conventional forward solution methods demanding consistent accuracy is highly dependent on the number of spatial nodes utilized by the core simulator, varying from superior to inferior performance as the number of nodes increases

  19. Nuclear co-generation desalination complex with VK-300 simplified boiling-water reactor

    International Nuclear Information System (INIS)

    Kuznetsov, Yury

    2008-01-01

    With regard for the global-scale development of desalination technologies and the stable growth demand for them, Russia also takes an active part in the development of these technologies. Two major aspects play a special role here: they are providing the desalination process with power and introducing new materials capable to make the production of fresh water cheaper and raise the technical reliability of desalination units. The report considers a simplified passive boiling water reactor VK-300 based Nuclear Desalination Complex (NDC) with multi-stage evaporation distillation desalination units (MED) with horizontal-tube film evaporators. This is the effective NDC structure allowing the use of turbine steam extractions for heat supply (200-400 Gcal/h) to the desalination system producing high-quality distillate. As it provides with thermal energy a desalination complex with the capacity of 300.000 m 3 /day, a nuclear plant consisting of two VK-300 power units allows production of distillate with the cost of 0.58 dollars/m 3 . In this case, the electricity supply to the power system is 357 MW(e). The electricity cost is 0.029 dollars/kWh. (author)

  20. Hierarchy level scheme for quasi-optimum fuel assembly loading in boiling water reactors

    International Nuclear Information System (INIS)

    Sekimizu, K.

    1978-01-01

    A quasi-optimum fuel assembly allocation scheme for boiling water reactors was proposed and confirmed. It is characteristic of the scheme that the criteria function is represented by fuel assembly allotment to fuel groups. For each fuel group, a required property is given beforehand, and fuel assemblies are allocated to the core to determine the group property as closely as possible. By using the scheme, a fuel assembly allocation is obtained that has a large cycle burnup within a restriction for the peak-to-average power ratio. Another allocation is obtained that results in a large burnup of discharged fuel using a different criteria function. However, it is impossible to obtain a strictly optimum solution for a given criteria function because of the vast number of possible fuel assembly allocations. The search range is reduced by adopting a two-step scheme. In the first step, an optimum allocation of fresh assemblies is searched for, based on proper criteria. Then, in the second step, without moving the fresh fuel assemblies, an allocation of reload fuel assemblies is determined that ascertains the required group property as closely as possible. Results of the numerical calculation show that the scheme is very useful for practical fuel assembly allocation

  1. Response to severe-accident policy statement: Boiling water reactor containment vulnerability assessment

    International Nuclear Information System (INIS)

    Gabor, J.R.; Burns, E.T.; Mairs, T.P.

    1989-01-01

    The U.S. Nuclear Regulatory Commission (NRC's) Severe-Accident Policy Statement, Safety Goal Policy Statement, Individual Plant Examination (IPE) Generic Letter, and Containment Performance Improvement (CPI) Program seek to characterize adequate containment performance. This paper describes a framework (developed through the cooperation of a number of utilities) within which the following can be accomplished: questions regarding plant-specific containment performance can be addressed; the impact of proposed plant modifications can be investigated and the results communicated to the NRC. The NRC is currently assessing the performance of all containment types under postulated severe-accident conditions. Issues have been raised by the NRC regarding containment performance because it is a final barrier protecting the public against the release of radionuclides under sever-accident conditions. In addition, there are several arenas where additional related issues may be raised, e.g., NUREG-1150 (final issue), IPE reviews by the NRC staff, recommended accident management strategies, accident management proposed generic letter, and the NRC generic evaluation of boiling water reactor. This paper presents the methodology developed in cooperation with a number of utilities to respond to the NRC initiatives requiring a plant-specific containment performance evaluation as part of the IPE process

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

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    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) [pt

  3. Analysis on special two-phase flow instabilities at full power natural circulation reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Y.J.; Yang, X.T.; Jiang, S.Y. [Tsinghua Univ., Beijing, BJ (China). Inst. of Nuclear Energy Technology

    2004-08-01

    The experiments were performed on the test loop HRTL-5, which simulates the geometry and system design of the 5 MW nuclear heating reactor. In a wide range of inlet sub-cooling, various flow instabilities were observed at p = 0.1 MPa and p = 1.5 MPa. Because of the different geometry design and operating conditions between the heating reactors and the boiling water reactors, the flow behavior presents great difference. Analysis shows: (1) under heating reactor conditions, sub-cooled boiling, condensation and void flashing are the fundamental thermodynamic processes; (2) sub-cooled boiling, condensation, void flashing and the compressibility of the steam space play an important role in the flow instabilities of the natural circulation system; (3) sub-cooled boiling instability, flashing instability, and flow excursion are the special instabilities at nuclear heating reactor conditions. (orig.)

  4. Benchmarking lattice physics data and methods for boiling water reactor analysis

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Edenius, M.; Harris, D.R.; Hebert, M.J.; Kapitz, D.M.; Pilat, E.E.; VerPlanck, D.M.

    1983-01-01

    The objective of the work reported was to verify the adequacy of lattice physics modeling for the analysis of the Vermont Yankee BWR using a multigroup, two-dimensional transport theory code. The BWR lattice physics methods have been benchmarked against reactor physics experiments, higher order calculations, and actual operating data

  5. Simulation of Boiling Flow Experiments Close to CHF with the NeptuneCFD Code

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2008-01-01

    A three-dimensional two-fluid code Neptune C FD has been validated against the Arizona State University (ASU) and DEBORA boiling flow experiments. Two-phase flow processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function has been implemented in the Neptune C FD code aiming to improve the prediction of flow parameters in the near-wall region. The capability of the code to predict the boiling flow regime close to critical heat flux (CHF) conditions has been verified on selected DEBORA experiments. To predict the onset of CHF regime, a simplified model based on the near-wall values of gas volume fraction was used. The results have shown that the code is able to predict the wall temperature increase and the sharp void fraction peak near the heated wall, which are characteristic phenomena for CHF conditions

  6. The Effective Convectivity Model for Simulation of Molten Metal Layer Heat Transfer in a Boiling Water Reactor Lower Head

    Directory of Open Access Journals (Sweden)

    Chi-Thanh Tran

    2013-01-01

    Full Text Available This paper is concerned with the development of approaches for assessment of core debris heat transfer and Control Rod Guide Tube (CRGT cooling effectiveness in case of a Boiling Water Reactor (BWR severe accident. We consider a hypothetical scenario with stratified (metal layer atop melt pool in the lower plenum. Effective Convectivity Model (ECM and Phase-Change ECM (PECM are developed for the modeling of molten metal layer heat transfer. The PECM model takes into account reduced convection heat transfer in mushy zone and compositional convection that enables simulations of noneutectic binary mixture solidification and melting. The ECM and PECM are (i validated against relevant experiments for both eutectic and noneutectic mixtures and (ii benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is then applied to the analysis of heat transfer in a stratified heterogeneous debris pool taking into account CRGT cooling. The PECM simulation results show apparent efficacy of the CRGT cooling which can be utilized as Severe Accident Management (SAM measure to protect the vessel wall from focusing effect caused by metallic layer.

  7. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  8. Development of an in-core fuel management tool for boiling water reactors

    International Nuclear Information System (INIS)

    Gilli, Luca; Wakker, Pieter H.; Elder, Brian R.

    2017-01-01

    The in-core fuel management of a nuclear reactor is a challenging task due to the virtually infinite number of loading patterns one could theoretically adopt. The ROSA (Reloading Optimization by Simulated Annealing) code is an optimization tool that has been successfully used in the last two decades to facilitate the core design of several Pressurized Water Reactors (PWRs). It is designed to perform a stochastic search for an optimal Loading Pattern (LP) using a simulated annealing algorithm. This corresponds to performing a depletion calculation for each one of the hundreds of thousands of unique LPs generated during the stochastic search. Therefore, speed is one of the most important requirements that the solvers used by the depletion tool must fulfill. ROSA's depletion analysis tool makes use of a particularly fast nodal method (known as the kernel method) for the evaluation of the power distribution associated with a particular LP. One of the strongest assumptions behind the kernel method is that the neutron migration length does not change considerably between the point where a neutron is generated and the point where the same neutron is absorbed. Although strong, this assumption is quite compatible with the neutronic characteristics of PWRs cores. In this paper we give an overview of the work done in order to develop a version of ROSA capable of performing the core design of Boiling Water Reactors (BWRs). We focus the discussion on the development of the depletion analysis tool by outlining the modifications of the kernel methods implemented in order to make the solver accurate for BWR cores. An improvement of the definition of the transport kernel is necessary to take the strong anisotropies characterizing the neutronic problem into account. These anisotropies arise due to the presence of strong changes in the moderator density and due to the presence of control blades. Furthermore, we are going to discuss how the boundary conditions are adopted by the

  9. Physical experiments. Reactor theory

    International Nuclear Information System (INIS)

    Korn, H.; Werle, H.; Bluhm, H.; Fieg, G.; Kappler, F.; Kuhn, D.; Lalovic, M.; Woll, D.; Kuefner, K.; Woznicki, Z.; Buckel, G.; Stehle, B.; Borgwaldt, H.

    1975-01-01

    The γ-spectrum in SNEAK 9C-1 and 9C-2 was measured by means of Si(Li) solid state detectors for verification of methods of shielding calculation. The blanket spectra turned out to be slightly harder than the spectra in the fissile zone; the plutonium spectra are slightly harder than the respective uranium spectra. This result is expected to be explained by studies to be carried out on the basis of a γ-transport program. For reactor theoretical calculations two 2-dimensional diffusion programs were compared with each other, and a 3-dimensional diffusion program was compared with a flux synthesis program. An improved source iteration scheme was drafted for the Karlsruhe Monte Carlo code. (orig.) [de

  10. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  11. Flow Boiling and Condensation Experiment (FBCE) for the International Space Station

    Science.gov (United States)

    Mudawar, Issam; O'Neill, Lucas; Hasan, Mohammad; Nahra, Henry; Hall, Nancy; Balasubramaniam, R.; Mackey, Jeffrey

    2016-01-01

    An effective means to reducing the size and weight of future space vehicles is to replace present mostly single-phase thermal management systems with two-phase counterparts. By capitalizing upon both latent and sensible heat of the coolant rather than sensible heat alone, two-phase thermal management systems can yield orders of magnitude enhancement in flow boiling and condensation heat transfer coefficients. Because the understanding of the influence of microgravity on two-phase flow and heat transfer is quite limited, there is an urgent need for a new experimental microgravity facility to enable investigators to perform long-duration flow boiling and condensation experiments in pursuit of reliable databases, correlations and models. This presentation will discuss recent progress in the development of the Flow Boiling and Condensation Experiment (FBCE) for the International Space Station (ISS) in collaboration between Purdue University and NASA Glenn Research Center. Emphasis will be placed on the design of the flow boiling module and on new flow boiling data that were measured in parabolic flight, along with extensive flow visualization of interfacial features at heat fluxes up to critical heat flux (CHF). Also discussed a theoretical model that will be shown to predict CHF with high accuracy.

  12. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Riley, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schroeder, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Aldrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maljovec, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Bie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pascucci, Valerio [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  13. Experience in utilizing research reactors in Yugoslavia

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

    1972-01-01

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  14. Sodium boiling detection in LMFBRs (Phase I). 5th quarterly technical progress report, July 1, 1975--October 31, 1975

    International Nuclear Information System (INIS)

    Albrecht, R.W.; McCormick, N.J.

    1975-01-01

    Progress summarized includes the design of a gamma heated subassembly for sodium boiling experiments and an experiment showing that neutronic noise and acoustic noise caused by sodium boiling are highly correlated in a wide frequency band about the bubble repetition frequency

  15. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Science.gov (United States)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  16. Analysis of the EJET boiling jet mixing experiments using the integrated fuel-coolant interaction code, IFCI

    International Nuclear Information System (INIS)

    Rightley, M.J.; Young, M.F.; Beck, D.F.

    1991-01-01

    In the event of a severe reactor accident leading to core melt, it is likely that molten fuel materials will come into contact with water, producing a molten fuel-coolant interaction (FCI). FCIs can occur for a variety of conditions in the core, the lower plenum, or in the reactor cavity. The nature of the FCIs that could occur ranges from benign static boiling, possibly including melt dispersion when the coherent melt mass is broken up on a time scale of 100's of milliseconds, to energetic steam explosions when the melt is finely fragmented on a time scale of milliseconds. Experimentation has revealed that scale-dependent processes occur in FCI's and that these dependencies are not understood. Attempts to model the process have generated several competing models. Unfortunately, the limited size and nature of the experimental database have made the choice of the correct model difficult. The integrated fuel-coolant interaction code, IFCI, was developed to provide a best estimate tool for FCIs, based on known physical laws and available experiments. The process of assessing the performance of IFCI involves comparing it to the different stages of FCI phenomena such as boiling jet breakup, detonation and products expansion. The NRC Program Molten Fuel-Coolant Interactions was initiated to perform this assessment against the current experimental data and other codes that have been developed to model FCIs. Upon completion of the assessment of the code, IFCI will be applied to reactor-scale simulations of lower plenum coarse mixing, steam and hydrogen production rates and steam explosion probabilities and their intensities

  17. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Knight, D.D.

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  19. Intercomparison of auto- and cross-power spectral density surveillance systems for sodium boiling detection in fast reactors

    International Nuclear Information System (INIS)

    Ehrhardt, J.

    1979-01-01

    Theoretical and experimental investigations on detection systems for small narrow-band components in noise signals were conducted. These detectionn systems are based on the continuous surveillance of the power spectral density for characteristic peaks. Detection sensitivity for auto- and cross-correlation measurements was computed for signals with normally distributed amplitudes in dependence of signal coherence. The derived detection criteria allowed the comparison of auto- and cross-power spectral density surveillance. Theoretical results were confirmed in a number of experimental parameter studies. Special theoretical investigations were done for the optimal detection of local sodium boiling in liquid-metal fast breeder reactors

  20. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  1. 10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

    Science.gov (United States)

    2010-01-01

    ...-operated valves. (4) Equipment seismic qualification methods. (5) Piping design acceptance criteria. (6... software qualification. (13) Self-test system design testing features and commitments. (14) Human factors... 10 Energy 2 2010-01-01 2010-01-01 false Design Certification Rule for the U.S. Advanced Boiling...

  2. Effects of torus wall flexibility on forces in the Mark I Boiling Water Reactor Pressure Suppression System. Part I

    International Nuclear Information System (INIS)

    Martin, R.W.; McCauley, E.W.

    1977-09-01

    The authors investigated the effects of torus wall flexibility in the pressure suppression system of a Mark I boiling water reactor (BWR) when the torus wall is subjected to hydrodynamic loadings. Using hypothetical models, they examined these flexibility effects under two hydrodynamic loading conditions: (1) a steam relief valve (SRV) discharge pulse, and (2) a loss-of-coolant accident (LOCA) chugging pulse. In the analyses of these events they used a recently developed two-dimensional finite element computer code. Taking the basic geometry and dimensions of the Monticello Mark I BWR nuclear power plant (in Monticello, Minnesota, U.S.A.), they assessed the effects of flexibility in the torus wall by changing values of the inside-diameter-to-wall-thickness ratio. Varying the torus wall thickness (t) with respect to the inside diameter (D) of the torus, they assigned values to the ratio D/t ranging from 0 (infinitely rigid) to 600 (highly flexible). In the case of a modeled steam relief valve (SRV) discharge pulse, they found the peak vertical reaction force on the torus was reduced from that of a rigid wall response by a factor of 3 for the most highly flexible, plant-simulated wall (D/t = 600). The reduction factor for a modeled loss-of-coolant accident (LOCA) chugging pulse was shown to be 1.5. The two-dimensional analyses employed overestimate these reduction factors but have provided, as intended, definition of the effect of torus boundary stiffness. In the work planned for FY79, improved modeling of the structure and of the source is expected to result in factors more directly applicable to actual pressure suppression systems

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

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

  4. Compound effects of operating parameters on burnup credit criticality analysis in boiling water reactor spent fuel assemblies

    Directory of Open Access Journals (Sweden)

    Shang-Chien Wu

    2018-02-01

    Full Text Available This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor (keff versus burnup (B are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE14 10 × 10 boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC-68 storage cask. The results revealed that the curves of keff versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of keff,Δk in some compound effects was not a summation of the all Δk resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of keff versus B for both single and compound effects.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. Simulation of sodium boiling experiments with THERMIT sodium version. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Huh, K.Y.

    1982-05-01

    Natural and forced convection experiments (SBTF and French) are simulated with the sodium version of the thermal-hydraulic computer code THERMIT. Simulation is done for the test section with the pressure-velocity boundary condition and subsequently extended to the whole loop. For the test section simulation, a steady-state and transient calculations are performed and compared with experimental data. For the loop simulation, two methods are used, a simulated 1-D loop and an actual 1-D loop. In the simulated 1-D loop analysis, the vapor density is increased by one hundred and two hundred times to avoid the code failure and the results still showed some of the important characteristics of the two-phase flow oscillation in a loop. A mathematical model is suggested for the two-phase flow oscillation. In the actual 1-D loop, only the single phase calculation was performed and turned out to be nearly the same as the simulated 1-D loop single phase results.

  9. Experience with spent fuel storage at research and test reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 July 1993

    International Nuclear Information System (INIS)

    1995-01-01

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work begins on appropriate solutions. To obtain an overall picture of the size and extent of these problems, an Advisory Group Meeting on Storage Experience with Spent Fuel from Research Reactors was convened in Vienna 5-8 July 1993, and attended by twelve participants and three observers representing thirteen different countries. These proceedings contain the country reports presented at the meeting. Refs, figs and tabs

  10. Crack propagation behavior of Ti-5Ta alloy in boiling nitric acid solution

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Kiuchi, Kiyoshi

    1999-05-01

    The crack propagation behavior of Ti-5Ta alloy both in boiling nitric acid solution and in air at room temperature has been investigated. The crack growth rate of Ti-5Ta alloy was measured as a function of the stress intensity factor range. After the tests, the fracture surface morphology was observed by a scanning electron microscope and the crystallographic orientation was examined by X-ray diffraction analysis. Difference in the crack growth behavior was not observed in both environments. The crack growth rate in boiling nitric acid solution was similar to that in air at room temperature. Moreover, the crystallographic orientation of Ti-5Ta alloy had little effect on the fatigue behavior, because this alloy does not have the susceptibility to SCC in nitric acid solution. (author)

  11. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  12. Conceptual design and safety characteristics of the natural circulation boiling water reactor HSBWR-600

    International Nuclear Information System (INIS)

    Naitoh, M.; Kataoka, Y.; Suzuki, H.; Sumida, I.; Horiuchi, T.; Akita, M.; Miki, M.

    1990-01-01

    The HSBWR (Hitachi Small BWR) with a rated capacity of 600 MW electricity has been conceptually designed. The components and systems are simplified by adopting natural circulation and the passive ECCS, and eliminating steam separators. The volume of the reactor building is about 50% of that for current BWRs with the same rated capacity, and the construction period is 32-36 months until commercial operation. The major safety systems are: (1) an accumulated water injection system as an ECCS; (2) an outer pool, which stands outside of the steel primary containment vessel, as a long term cooling system after LOCAs; and (3) a steam driven reactor core isolation cooling system for high pressure water injection. The grace period is one day for core cooling and 3 days for the containment vessel heat removal. The infinite grace period for core cooling is also available as an option. LOCA analysis showed that the core will always be covered by a two-phase mixture, resulting in no core heat-up. The fundamental experiments and analyses showed sufficient capability of the outer pool for long term heat removal. (author). 12 refs, 17 figs, 3 tabs

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

    International Nuclear Information System (INIS)

    Brettschuh, Werner

    2006-01-01

    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

  14. Experimental Study on Boiling Regime During Quenching Process in Heated Rod Bundle Queen

    International Nuclear Information System (INIS)

    J, Mulya; Antariksawan, A.R.; PW, Joko; S, Edy; H, Khairul; H, Ismu; Kiswanta; Giarno

    2003-01-01

    Following loss-of-coolant accident in light water reactor, the emergency core cooling must be injected. During flooding the core, the fuel cladding quenching occurred. The fuel quenching velocity is key factor for reactor safety. Various parameter influence the quenching velocity. It can also be related to the boiling regime change during transient. Current experimental study is performed to observe and apprehend boiling regime during quenching process and to measure its velocity. Experiment is conducted using Queen heated rod bundle. The quenching occurred from bottom flooding with flow rate of 0.0417 kg/s. The initial temperature of heated rod varies from 334 o C at zero point and 499 o C at top of heated zone. The visual observation method and rod surface temperature measurements is used to discus the change of boiling regime and quench front velocity. From the observation, it is obvious that at a one defined point, the boiling regime change from film boiling to single phase convection. On the other hand, the quench front velocity was affected by surface temperature and boiling regime. At the heated zone and at the beginning of quench, the quench front velocity was relatively low. While the surface temperature decreases, the quench front velocity was increase until all vapor film collapse. The average quench front velocity is about 11.5 mm/s

  15. The study of flow resistance in nuclear reactor Maria under coolant boiling condition

    International Nuclear Information System (INIS)

    Czerski, P.

    1999-01-01

    This study describes an analysis of experiments carried out in the WIW-300 installation located in the Institute of Atomic Energy (Swierk, Poland). The flow, simulated in the annular gap of test section, was similar to the flow in Maria reactor fuel channel. Experimental character of the work lead to the conclusions related to the physical nature of the hydrodynamic phenomena investigated as well as to the practical aspects of future research. A hypothesis defining a cause of pressure changes was formulated and specific problems related to the mathematical model were defined. The analysis shows that hydrodynamic phenomena studies are of basic significance for the prediction of burnout effects and that heat exchange is very often determined by local phenomena. All described observations are the base for further research on thermodynamic aspects of investigated phenomena. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Trianti, Nuri, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Su' ud, Zaki, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id; Arif, Idam, E-mail: nuri.trianti@gmail.com, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Riyana, EkaSapta [Nuclear Energy Regulatory Agency (BAPETEN) (Indonesia)

    2014-09-30

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

  17. Analysis of flow boiling heat transfer in narrow annular gaps applying the design of experiments method

    Directory of Open Access Journals (Sweden)

    Gunar Boye

    2015-06-01

    Full Text Available The axial heat transfer coefficient during flow boiling of n-hexane was measured using infrared thermography to determine the axial wall temperature in three geometrically similar annular gaps with different widths (s = 1.5 mm, s = 1 mm, s = 0.5 mm. During the design and evaluation process, the methods of statistical experimental design were applied. The following factors/parameters were varied: the heat flux q · = 30 − 190 kW / m 2 , the mass flux m · = 30 − 700 kg / m 2 s , the vapor quality x · = 0 . 2 − 0 . 7 , and the subcooled inlet temperature T U = 20 − 60 K . The test sections with gap widths of s = 1.5 mm and s = 1 mm had very similar heat transfer characteristics. The heat transfer coefficient increases significantly in the range of subcooled boiling, and after reaching a maximum at the transition to the saturated flow boiling, it drops almost monotonically with increasing vapor quality. With a gap width of 0.5 mm, however, the heat transfer coefficient in the range of saturated flow boiling first has a downward trend and then increases at higher vapor qualities. For each test section, two correlations between the heat transfer coefficient and the operating parameters have been created. The comparison also shows a clear trend of an increasing heat transfer coefficient with increasing heat flux for test sections s = 1.5 mm and s = 1.0 mm, but with increasing vapor quality, this trend is reversed for test section 0.5 mm.

  18. Instabilities in parallel channel of forced-convection boiling upflow system, 5

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Aoki, Shigebumi; Inoue, Akira

    1983-01-01

    The density wave instability in a parallel boiling channel system heated electrically has been studied experimentally and analytically by the authors. In our country, the steam generator for LMFBR has been investigated with Power Reactor and Nuclear Fuel Development Corp. as the central figure for its development, and many results of this instability were reported. Their results were different from our ones as regard to the governing factor of the period of flow oscillation in the unstable region and to the effect of the slip ratio on the stability in analysis. A new linear analytical model is proposed in this paper and the analytical results are compared with ones of two-phase analyses based on the same linear method as this model. Subsequently, the effect of the slip ratio on the stability is studied analytically by this model. The parallel boiling channel system is studied experimentally and analytically, using Freon-113 as test fluid heated by hot water as simulation of the SG for LMFBR. The governing factor of the period of flow oscillation is made clear. (author)

  19. Lessons from early experience in reactor development

    International Nuclear Information System (INIS)

    Allen, W.

    1976-09-01

    This paper deals with several issues in U.S. reactor development and demonstration experience. The focus is on the period between 1946 and 1963 during which the Atomic Energy Commission (AEC) guided early reactor research and development (R and D) and conducted the Power Reactor Demonstration Program

  20. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Energy Technology Data Exchange (ETDEWEB)

    Rubio, Rafael, E-mail: rrubio@iberdrola.es [Iberdrola Generación Nuclear S.A., Madrid (Spain); Jimenez, Gonzalo [Universidad Politécnica de Madrid (Spain)

    2014-08-15

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

  1. Checking technical measurements on climatic data during sand blasting and spraying work in the condensation chamber of the boiling water reactor Gundremmingen

    International Nuclear Information System (INIS)

    Rausch, D.; Unte, U.

    1986-01-01

    During sand blasting and spraying work in the condensation chambers of boiling water reactors prescribed climatic data must be adhered to. For this purpose temporary air conditioners are used. The technical measurement examination here should provide information as to whether the air conditioners used were to fulfill the parameter curve specifications. (orig.) [de

  2. The construction and initial startup of an onfloated pulver resin filter equipment for condensate treatment at a boiling water reactor

    International Nuclear Information System (INIS)

    Gruenewale, D.; Wieland, G.

    1978-01-01

    The treatment of condensate at the boiling water reactors is owing to the corrosive and splitting products, as well as condenser leakages obligatory. The various possibilities of condensate treatment flow diagrams are outlined. Then the procedure of the pulver resin onfloating and the functioning phases of the respective equipment are shown by the means of illustration materials. After the startup of the equipment the condensate make-up equipment presented a statisfactory function within a short time. After putting the power plant into service the condensate make-up equipment delivered an adequately clean condenste. The clefts arising under certain operational circumstances caused a weaker quality at the onfloating, but they could be eliminated in a short time. The comparison of costs with the mix-bed filters points out to the advantages of the onfloated filters until 500 l/h condenser leakage. (author)

  3. Method for optimum determination of adjustable parameters in the boiling water reactor core simulator using operating data on flux distribution

    International Nuclear Information System (INIS)

    Kiguchi, T.; Kawai, T.

    1975-01-01

    A method has been developed to optimally and automatically determine the adjustable parameters of the boiling water reactor three-dimensional core simulator FLARE. The steepest gradient method is adopted for the optimization. The parameters are adjusted to best fit the operating data on power distribution measured by traversing in-core probes (TIP). The average error in the calculated TIP readings normalized by the core average is 0.053 at the rated power. The k-infinity correction term has also been derived theoretically to reduce the relatively large error in the calculated TIP readings near the tips of control rods, which is induced by the coarseness of mesh points. By introducing this correction, the average error decreases to 0.047. The void-quality relation is recognized as a function of coolant flow rate. The relation is estimated to fit the measured distributions of TIP reading at the partial power states

  4. In-core power sharing and fuel requirement study for a decommissioning Boiling Water Reactor using the linear reactivity model

    International Nuclear Information System (INIS)

    Chen, Chung-Yuan; Tung, Wu-Hsiung; Yaur, Shung-Jung; Kuo, Weng-Sheng

    2014-01-01

    Highlights: • Linear reactivity model (LRM) was modified and applied to Boiling Water Reactor. • The power sharing and fuel requirement study of the last cycle and two cycles before decommissioning was implemented. • The loading pattern design concept for the cycles before decommissioning is carried out. - Abstract: A study of in-core power sharing and fuel requirement for a decommissioning BWR (Boiling Water Reactor) was carried out using the linear reactivity model (LRM). The power sharing of each fuel batch was taken as an independent variable, and the related parameters were set and modified to simulate actual cases. Optimizations of the last cycle and two cycles before decommissioning were both implemented; in the last-one-cycle optimization, a single cycle optimization was carried out with different upper limits of fuel batch power, whereas, in the two-cycle optimization, two cycles were optimized with different cycle lengths, along with two different optimization approaches which are the simultaneous optimization of two cycles (MO) and two successive single-cycle optimizations (SO). The results of the last-one-cycle optimization show that it is better to increase the fresh fuel power and decrease the thrice-burnt fuel power as much as possible. It also shows that relaxing the power limit is good to the fresh fuel requirement which will be reduced under lower power limit. On the other hand, the results of the last-two-cycle (cycle N-1 and N) optimization show that the MO is better than SO, and the power of fresh fuel batch should be decreased in cycle N-1 to save its energy for the next cycle. The results of the single-cycle optimization are found to be the same as that in cycle N of the multi-cycle optimization. Besides that, under the same total energy requirement of two cycles, a long-short distribution of cycle length design can save more fresh fuel

  5. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  6. 1/5-scale experiment of a Mark I boiling-water reactor pressure-suppression system under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    Experimental results show the sensitivity of hydrodynamically generated vertical loads to changes in the drywell pressurization rate, downcomer submergence, and vent-line loss coefficient. Insignificant effects on peak vertical loads were observed when the vent-line loss was varied. Peak vertical loads can be reduced by adding initial drywell overpressure so that the downcomers are partly cleared of water. Spatial variation of pressure at about the time of vent clearing is seen in comparisons of data from locations along the axis of the toroidal wetwell

  7. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2014-07-01

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. The paper describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motorgenerator set.

  8. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    International Nuclear Information System (INIS)

    Loenhout, Gerard van; Hurni, Juerg

    2014-01-01

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. The paper describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motorgenerator set.

  9. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  10. Performance Evaluation of the International Space Station Flow Boiling and Condensation Experiment (FBCE) Test Facility

    Science.gov (United States)

    Hasan, Mohammad; Balasubramaniam, R.; Nahra, Henry; Mackey, Jeff; Hall, Nancy; Frankenfield, Bruce; Harpster, George; May, Rochelle; Mudawar, Issam; Kharangate, Chirag R.; hide

    2016-01-01

    A ground-based experimental facility to perform flow boiling and condensation experiments is built in support of the development of the long duration Flow Boiling and Condensation Experiment (FBCE) destined for operation on board of the International Space Station (ISS) Fluid Integrated Rack (FIR). We performed tests with the condensation test module oriented horizontally and vertically. Using FC-72 as the test fluid and water as the cooling fluid, we evaluated the operational characteristics of the condensation module and generated ground based data encompassing the range of parameters of interest to the condensation experiment to be performed on the ISS. During this testing, we also evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, and the heat loss from different components. In this presentation, we discuss representative results of performance testing of the FBCE flow loop. These results will be used in the refinement of the flight system design and build-up of the FBCE which is scheduled for flight in 2019.

  11. Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

    International Nuclear Information System (INIS)

    Uegata, T.; Saji, E.; Tanaka, H.

    1993-01-01

    Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

  12. Calculation model for predicting concentrations of radioactive corrostion products in the primary coolant of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, S.; Kikuchi, M.; Asakura, Y.; Yusa, H.; Ohsumi, K.

    1978-01-01

    A calculation model was developed to predict the shutdown dose rate around the recirculation pipes and their components in boiling water reactors (BWRs) by simulating the corrosion product transport in primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating each concentration using the following considerations: (1) Insoluble cobalt (designated as crud cobalt is deposited directly on the fuel surface, while soluble cobalt (designated as ionic cobalt) is adsorbed on iron oxide deposits on the fuel surface. (2) Cobalt-60 activated on the fuel surface is dissolved in the water in an ionic form, and some is released with iron oxide as crud. The model can follow the reduction of 60 Co in the primary cooling water caused by the control of the iron feed rate into the reactor, which decreases the iron oxide deposits on the fuel surface and then reduces the cobalt adsorption rate. The calculated results agree satisfactorily with the measurements in several BWR plants

  13. Recent U.S. reactor operating experience

    International Nuclear Information System (INIS)

    Stello, V. Jr.

    1977-01-01

    A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

  14. Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

    International Nuclear Information System (INIS)

    Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

    1988-01-01

    In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

  15. Study on dual plant concept for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    Sato, Takashi; Oikawa, Hirohide

    1999-01-01

    The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

  16. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Energy Technology Data Exchange (ETDEWEB)

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2015-05-15

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  17. New contact boiling experiments to evaluate Calandria tube strain acceptance criteria

    Energy Technology Data Exchange (ETDEWEB)

    El-Hawary, M.; Szymanski, J.; Tanase, A.; Delja, A.; Oussoren, A., E-mail: Magdy.El-Hawary@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Comission, Ottawa, ON (Canada); Neal, P. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    The Canadian Nuclear Safety Commission(CNSC) has contracted the Canadian Nuclear Laboratories(CNL) to conduct additional Contact Boiling (CB) experiments with the main objective of evaluating the acceptance criterion of CalandriaTube (CT) strain limit of 2%, proposed by the industry for fuel channel integrity assessments. The test conditions are selected using analytical tools and guidance from existing CANDU Owners Group (COG) test results, so as to lead to CT strain close to this value. The experiments will also be used to evaluate the CT quench temperature correlation proposed. This paper presents conditions selected for the first three experiments, their most important results and their preliminary analysis, with a focus on the test which produced CT strain in excess of 2%. (author)

  18. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

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

    Science.gov (United States)

    2012-01-20

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

  20. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  1. Reconstruction of local heat fluxes in pool boiling experiments using the entire heater geometry

    Energy Technology Data Exchange (ETDEWEB)

    Heng, Y.; Mhamdi, A.; Marquardt, W. [RWTH Aachen University, Aachen (Germany). AVT-Process Systems Engineering; Buchholz, M.; Auracher, H. [Berlin University of Technology (Germany). Inst. for Energy Engineering

    2009-07-01

    In this work, we consider the reconstruction of local boiling heat fluxes from high resolution transient temperature measurements inside the heater obtained during experiments performed at T U Berlin. In our previous work, a very small 3D domain surrounding the micro thermocouples at the center of a test heater has been considered. The unknown lateral boundary conditions have been set to zero, due to the lack of better knowledge. This geometry and the related assumptions have been chosen, due to the computational limitations. In the present study, we address for the first time this problem over the entire test heater. This has been only possible by improving the computational efficiency and using a suitable non-uniform discretization strategy. The boundary conditions in this study are well-defined at the boundaries where no boiling occurs. We formulate the heat flux estimation as a three-dimensional transient inverse heat conduction problem (IHCP). The solution of this ill-posed problem is obtained by applying an iterative regularization strategy, which is a combination of the method of conjugate gradients for the normal equation and the discrepancy principle. The obtained results are similar to those obtained in our previous work. The estimates are, however, much better in this work, since we not only recover the dynamics of the signal, but also largely avoid the negative heat fluxes, which we observed using the much smaller region. (author)

  2. Influence of Boiling Duration of GCSB-5 on Index Compound Content and Antioxidative and Anti-inflammatory Activity.

    Science.gov (United States)

    Lee, In-Hee; Chung, Hwa-Jin; Shin, Joon-Shik; Ha, In-Hyuk; Kim, Me-Riong; Koh, Wonil; Lee, Jinho

    2017-01-01

    GCSB-5, an herbal drug composition with an anti-inflammatory effect, is prepared by boiling, which is the most common herbal extraction method in traditional Korean medicine. Several parameters are involved in the process, i.e., extractant type, herb-to-extractant ratio, extraction temperature and pressure, and total boiling time. The aim of this study was to examine the influence of boiling time on index compound amount and the antioxidative and anti-inflammatory activities of GCSB-5. Different samples of GCSB-5 were obtained by decocting for 30, 60, 90, 120, 150, and 240 min. Each sample was tested for hydrogen ion concentration (pH), total soluble solid content (TSSC), marker compound profiles, and antioxidative and anti-inflammatory activity. pH was found to decrease while TSSC increased with extended decoction. Marker compound contents for GCSB-5 (acanthoside D for Acanthopanax sessiliflorus Seem, 20-hydroxyecdysone for Achyranthes japonica Nakai, and pinoresinol diglucoside for Eucommia ulmoides Oliver) remained relatively constant regardless of the length of boiling. Total D-glucose amount increased with longer boiling. The antioxidative and anti-inflammatory potentials of GCSB-5 were not substantially affected by decoction duration. Biological characteristics and marker compound content of GCSB-5 were not altered significantly in prolonged boiling. Longer boiling duration of GCSB-5 did not increase yield in a time-dependent manner, but yields of 210 and 240 min samples were significantly higherHydrogen ion concentration of GCSB-5 samples decreased while total soluble solid content and D-glucose concentration levels increased with boiling durationAlthough concentrations of some index compounds increased with extended boiling duration of GCSB-5, increase was small and not in a direct proportional relationshipAntioxidative and anti-inflammatory properties of GCSB-5 were not substantially affected by decoction duration. Abbreviations used: CAM: Complementary

  3. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  4. Reactor physics experiment plan using TCA

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX-fuel rods used in the experiments is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

  5. Reactor Neutrino Experiments: Present and Future

    Science.gov (United States)

    Wen, L. J.; Cao, J.; Wang, Y. F.

    2017-10-01

    Reactor neutrinos have been an important tool for both discovery and precision measurement in the history of neutrino studies. Since the first generation of reactor neutrino experiments in the 1950s, the detector technology has advanced greatly. New ideas, new knowledge, and modern software have also enhanced the power of the experiments. The current reactor neutrino experiments, Daya Bay, Double Chooz, and RENO, have led neutrino physics into the precision era. In this article, we review these developments and advances, address the key issues in designing a state-of-the-art reactor neutrino experiment, and explain how the challenging requirements of determining the neutrino mass hierarchy with the next-generation experiment JUNO could be realized in the near future.

  6. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  7. Improvement of boiling heat transfer by radiation induced boiling enhancement

    International Nuclear Information System (INIS)

    Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

    2003-01-01

    For nuclear reactor systems, the critical heat flux (CHF) data is very important because it limits reactor efficiency. Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, surface wettability in a high-temperature environment and critical heat flux (CHF) of metal oxides irradiated by gamma rays were investigated. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure. The heating test section made of titanium was 0.2 mm in thickness, 3 mm in height, and 60 mm in length. Oxidation of the surface was carried out by plasma jetting for 40 seconds. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. The CHF of oxidized titanium was improved up to 100 percent after 800 kGy 60 Co gamma ray irradiation. We call this effect Radiation Induced Boiling Enhancement (RIBE). Before we conducted the CHF experiment, contact angles of the test pieces were measured to show the relationship between wettability and CHF. The CHF in the present experiment increases will surface wettability in the same manner as shown by Liaw and Dhir's results. (author)

  8. A parametric study of the steady-state operational characteristics of the Ohio State University natural circulation indirect-cycle, inherently safe boiling water reactor

    International Nuclear Information System (INIS)

    Aybar, H.S.

    1995-01-01

    The Ohio State University Inherently Safe Reactor (OSU-ISR) is a conceptual design for a 340-MW(electric) [1,000-MW(thermal)], natural circulation, indirect-cycle, small boiling water reactor. All the OSU-ISR primary loop components are housed within a prestressed concrete reactor vessel (PCRV). The OSU-ISR performance has been investigated as a function of several design parameters in an attempt to better understand the interdependency among the system variables and hence to establish a knowledge base for the refinement of the conceptual design. The computational tool used in the study is a Dynamic Simulation for Nuclear Power Plants (DSNP) code whose predictions for the steady-state OSU-ISR performance compare favorably with RELAP5/MOD3 results for most of the operational characteristics of interest. The results show that (a) the key quantity that governs the OSU-ISR steady-state performance is the pressure difference between the primary and the secondary loops, (b) the magnitude of water-level swell (which occurs due to void formation in the core during operation and which affects the size of the steam separators that need to be used) can be more effectively controlled by varying the PCRV water level at cold shutdown rather than by varying the internal PCRV dimensions, (c) turbine inlet steam quality can be controlled without substantially affecting the other operational parameters by varying the secondary mass flow rate, and (d) the PCRV pressure and core exit steam quality are most sensitive to changes in the secondary loop pressure. The results also show that if there is a large drop in the secondary loop pressure (e.g., due to a steam line break), then although this pressure drop may induce a large drop in the PCRV pressure, the core flow, and hence core cooling capability, will not be appreciably affected

  9. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  10. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 1. Main report. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. Sloshing of water in annular pressure-suppression pool of boiling water reactors under earthquake ground motions

    International Nuclear Information System (INIS)

    Aslam, M.; Godden, W.G.; Scalise, D.T.

    1979-10-01

    This report presents an analytical investigation of the sloshing response of water in annular-circular as well as simple-circular tanks under horizontal earthquake ground motions, and the results are verified with tests. This study was motivated because of the use of annular tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 120 ft inside and outside diameters and a water depth of 20 ft. The analysis was based upon potential flow theory and a computer program was written to obtain time-history plots of sloshing displacements of water and the dynamic pressures. Tests were carried out on 1/80th and 1/15th scale models under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding the natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found for relatively small displacements. The computer program gave satisfactory results as long as the maximum water surface displacements were less than 30 in., which is roughly the value obtained under full intensity of El Centro earthquake

  13. Dynamic PIV measurement of the effect of sound waves in the upper plenum of the boiling water reactor

    International Nuclear Information System (INIS)

    Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

    2008-01-01

    In recent years, power uprating of boiling power reactors has been conducted at several existing power plants in order to improve plant economy. In one power uprated plant (117.8% uprate) in the United States, steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound waves into the steam-dome. The resonance among the structure, the flow, and the pressure fluctuation resulted in the breakages. In order to clarify the basic mechanism of the resonance, previous studies were performed by conducting a point measurement of the pressure and a phase averaged measurement of the flow, although detecting the interaction among the structure, the flow, and the pressure fluctuation by the conventional method was difficult. In a preliminary study, a dynamic Particle Image Velocimetry (PIV) system was used to investigate the effect of sound on the flow. A dynamic PIV system is the newest entrant to the field of fluid flow measurement. Its paramount advantage is the instantaneous global evaluation of conditions over a plane extended across the entire velocity field. Using the dynamic PIV system, the influence of sound waves on the flow field was measured. As a result, when two speakers were placed diagonally and sound waves were presented in the same phase, vertical motion was strongly observed compared to horizontal motion. (author)

  14. Failure analysis of the standby liquid control system for a boiling water reactor with fuzzy cognitive maps

    International Nuclear Information System (INIS)

    Nunez-Carrera, Alejandro; Espinosa-Paredes, Gilberto; Cruz-Esteban, Hugo

    2011-01-01

    Highlights: → FCMs are proposed in order to determine failure modes in systems and equipment in BWRs. → A simplified model is compared with the fault tree analysis technique. → Five case scenarios are studied in order to test the performance of the method. → The proposed method shows consistency with the traditional fault tree technique. - Abstract: A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, in order to see that both techniques show similar results even if the approaches are different.

  15. Experiments with the SUR 100 training reactor

    International Nuclear Information System (INIS)

    Milicic, B.

    1984-06-01

    This paper contains a compilation of various experiments using the SUR - 100 reactor for training purposes, which have been widly proved in practical work at the School for Nuclear Technology of the Karlsruhe Research Center. (orig.) [de

  16. Improved control rod drive handling equipment for BWRs [boiling-water reactors]: Final report

    International Nuclear Information System (INIS)

    Turner, A.P.L.; Gorman, J.A.

    1987-08-01

    Improved equipment for removing and replacing control rod drives (CRDs) in BWR plants has been designed, built and tested. Control rod drives must be removed from the reactor periodically for servicing. Removal and replacement of CRDs using equipment originally supplied with the plant has long been recognized as one of the more difficult and highest radiation exposure maintenance operations that must be performed at BWR plants. The improved equipment was used for the first time at Quad Cities, Unit 2, during a Fall 1986 outage. The trial of the equipment was highly successful, and it was shown that the new equipment significantly improves CRD handling operations. The new equipment significantly simplifies the sequence of operations required to lower a CRD from its housing, upend it to a horizontal orientation, and transport it out of the reactor containment. All operations of the new equipment are performed from the undervessel equipment handling platform, thus, eliminating the requirement for a person to work on the lower level of the undervessel gallery which is often highly contaminated. Typically, one less person is required to operate the equipment than were used with the older equipment. The new equipment incorporates a number of redundant and fail safe features that improve operations and reduce the chances for accidents

  17. Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

    International Nuclear Information System (INIS)

    Dehollander, W.R.

    1979-01-01

    A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

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

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

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

  19. Summary of the First Workshop on OECD/NRC boiling water reactor turbine trip benchmark

    International Nuclear Information System (INIS)

    2000-11-01

    The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system. A BWR TT benchmark exercise, based on a well-defined problem with complete set of input specifications and reference experimental data, has been proposed for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. The benchmark consists of three separate exercises: Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data). Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation. Exercise 3 - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling. This first workshop was focused on technical issues connected with the first draft of

  20. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  1. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    International Nuclear Information System (INIS)

    Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

    2016-01-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  2. Analysis of the rotation accident of assemblies in boiling water reactors

    International Nuclear Information System (INIS)

    Becerril-Gonzalez M, J. J.; Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R.

    2012-10-01

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

  3. Boiling in porous media

    International Nuclear Information System (INIS)

    1998-01-01

    This conference day of the French society of thermal engineers was devoted to the analysis of heat transfers and fluid flows during boiling phenomena in porous media. This book of proceedings comprises 8 communications entitled: 'boiling in porous medium: effect of natural convection in the liquid zone'; 'numerical modeling of boiling in porous media using a 'dual-fluid' approach: asymmetrical characteristic of the phenomenon'; 'boiling during fluid flow in an induction heated porous column'; 'cooling of corium fragment beds during a severe accident. State of the art and the SILFIDE experimental project'; 'state of knowledge about the cooling of a particulates bed during a reactor accident'; 'mass transfer analysis inside a concrete slab during fire resistance tests'; 'heat transfers and boiling in porous media. Experimental analysis and modeling'; 'concrete in accidental situation - influence of boundary conditions (thermal, hydric) - case studies'. (J.S.)

  4. Results of acoustic measurements with an electric boiling generator at KNK II

    International Nuclear Information System (INIS)

    Aberle, J.

    1987-08-01

    With regard to an integral core surveillance in sodium-cooled breeder reactors acoustic measurement techniques are under development. To determine experimentally the acoustic transfer function of a reactor core and to demonstrate the detectability of local sodium boiling, experiments with a so-called Boiling Generator were carried out in the KNK II reactor. The main part of this Boiling Generator was an electrically heated pin bundle which was equipped with a local blockage to obtain cooling disturbances. In this report the results of the acoustic measurements carried out with the Boiling Generator are presented. Main topic of the evaluation is the determination of the acoustic transfer function between the core and the upper sodium plenum. The signal conditioning necessary prior to this investigation is also explained. Great effort was required to suppress electrical disturbances which superimposed the acoustic signals and could not be eliminated by the hardware during the experiments. Finally, the detectability of local boiling using acoustic measurements is considered

  5. The double chooz reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Botella, I Gil [CIEMAT, Basic Research Department, Avenida Complutense, 22, 28040 Madrid (Spain)], E-mail: ines.gil@ciemat.es

    2008-05-15

    The Double Chooz reactor neutrino experiment will be the next detector to search for a non vanishing {theta}{sub 13} mixing angle with unprecedented sensitivity, which might open the way to unveiling CP violation in the leptonic sector. The measurement of this angle will be based in a precise comparison of the antineutrino spectrum at two identical detectors located at different distances from the Chooz nuclear reactor cores in France. Double Chooz is particularly attractive because of its capability to measure sin{sup 2} (2{theta}{sub 13}) to 3{sigma} if sin{sup 2}(2{theta}{sub 13}) > 0.05 or to exclude sin{sup 2}(2{theta}{sub 13}) down to 0.03 at 90% C.L. for {delta}m{sup 2} = 2.5 x 10{sup -3} eV{sup 2} in three years of data taking with both detectors. The construction of the far detector starts in 2008 and the first neutrino results are expected in 2009. The current status of the experiment, its physics potential and design and expected performance of the detector are reviewed.

  6. Tritium experience in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Brooks, J.N. [Argonne National Lab., IL (United States); Hogan, J. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  7. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  8. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  9. Critical experiments on minimal-content gadolinia for above-5wt% enrichment fuels in Toshiba NCA

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Watanabe, Shouichi; Yoshioka, Kenichi; Mitsuhashi, Ishi; Kumanomido, Hironori; Sugahara, Satoshi; Hiraiwa, Kouji

    2009-01-01

    A concept of 'minimal-content gadolinia' with a content of less than several hundred ppm mixed in the 'above-5wt% enrichment UO 2 fuel' for super high burnup is proposed for ensuring the criticality safety in the UO 2 fuel fabrication facility for light water reactors (LWRs) without increase in investment cost. Required gadolinia contents calculated were from 53 to 305 ppm for enrichments of UO 2 powders for boiling water reactor (BWR) fuel from 6 to 10 wt%. It is expected that the minimal-content gadolinia yields an acceptable reactivity suppression at the beginning of operating cycle and no reactivity penalty at the end of operating cycle due to no residual gadolinium. A series of critical experiments were carried out in the Toshiba Nuclear Critical Assembly (NCA). Reactivity effects of the gadolinia were measured to clarify the nuclear characteristics, and the measured values and the calculated values agreed within 5%. (author)

  10. Optimization of operation schemes in boiling water reactors using neural networks

    International Nuclear Information System (INIS)

    Ortiz S, J. J.; Castillo M, A.; Pelta, D. A.

    2012-10-01

    In previous works were presented the results of a recurrent neural network to find the best combination of several groups of fuel cells, fuel load and control bars patterns. These solution groups to each problem of Fuel Management were previously optimized by diverse optimization techniques. The neural network chooses the partial solutions so the combination of them, correspond to a good configuration of the reactor according to a function objective. The values of the involved variables in this objective function are obtained through the simulation of the combination of partial solutions by means of Simulate-3. In the present work, a multilayer neural network that learned how to predict some results of Simulate-3 was used so was possible to substitute it in the objective function for the neural network and to accelerate the response time of the whole system of this way. The preliminary results shown in this work are encouraging to continue carrying out efforts in this sense and to improve the response quality of the system. (Author)

  11. Boiling eXperiment Facility (BXF) Fluid Toxicity Technical Interchange Meeting (TIM) with the Payload Safety Review Panel (PSRP)

    Science.gov (United States)

    Sheredy, William A.

    2012-01-01

    A Technical Interchange meeting was held between the payload developers for the Boiling eXperiment Facility (BXF) and the NASA Safety Review Panel concerning operational anomaly that resulted in overheating one of the fluid heaters, shorted a 24VDC power supply and generated Perfluoroisobutylene (PFiB) from Perfluorohexane.

  12. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  13. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  14. Results of the Nucifer reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

    2016-07-01

    Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

  15. Operational experience of the Marcoule reactors

    International Nuclear Information System (INIS)

    Conte, F.

    1963-01-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

  16. A theoretical model for coupled neutronic-thermohydraulic out-of-phase oscillations in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Bragt, D.D.B. van.

    1995-10-01

    A theoretical model for out-of-phase power oscillations in BWRs is proposed. This model describes the dynamic behavior of the neutronic and thermohydraulic subsystems during out-of-phase oscillations, and the coupling of these subsystems via the fuel temperature dynamics and void- and Doppler feedback effects. The zero-power neutron kinetics of the out-of-phase flux density mode is derived by expanding the (time- and space-dependent) neutron flux density in the static solutions of the neutron transport equation. This procedure yields the modal point-kinetic equations for the (first-harmonic) out-of-phase mode. The fuel temperature dynamics is described by a lumped parameter first-order process, characterized by a typical fuel time constant. Using the quasistatic approach, the basic equations of the channel thermohydraulics are derived from the conservation laws of mass and energy and the momentum equation. The momentum equation is coupled with the appropriate boundary condition (constant core pressure drop) for out-of phase oscillations. This procedure yields a set of nonlinear equations describing the dynamic behavior of the boiling boundary, void fraction and mass flux density in the cooling channel. A frequency-domain parametric study confirms that if the out-of-phase mode has a more negative subcriticality, reactor stability increases. On the other hand, a more negative void reactivity coefficient has a destabilizing effect. Besides these two parameters, the fuel time constant was found to be an important parameter determining stability. Where possible, the linearized equations describing the channel thermohydraulics were compare with exact solutions of the governing partial-differential channel equations. This comparison shows that in the frequency range of interest, discrepancies between the proposed quasi-static model and more complicated exact solutions are to be expected. (orig.)

  17. Present and future oscillation experiments at reactors

    International Nuclear Information System (INIS)

    Mikaehlyan, L.A.

    2001-01-01

    A report is presented on recent progress and developments (since the NANP'99 Conference) in the current and future long baseline (∼100 - 800 km) oscillation experiments at reactors. These experiments, under certain assumptions, can fully reconstruct the internal mass structure of the electron neutrino and provide a laboratory test of solar and atmospheric neutrino problems

  18. State of the art report on boiling water reactor stability (SOAR on BWRs)

    International Nuclear Information System (INIS)

    1997-01-01

    Starting issues of this SOAR are BWR plant descriptions including peculiarities relevant to stability and the manifestation of instabilities during operation. The report continues with the characterization of instabilities from various experiments, the features and the capabilities of relevant codes and models, BWR core instrumentation and control, the stability behaviour of operating BWR plants and the regulatory approach to the stability issue. The main conclusion is that the BWR stability should not be considered as a safety issue; however R and D in specific areas is recommended

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

    International Nuclear Information System (INIS)

    Koncar, B.; Krepper, E.; Egorov, Y.

    2005-01-01

    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)

  20. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fuentes C, P

    2003-07-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O{sub 2}; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  1. Return to nucleate boiling

    International Nuclear Information System (INIS)

    Shumway, R.W.

    1985-01-01

    This paper presents a collection of TMIN (temperature of return to nucleate boiling) correlations, evaluates them under several conditions, and compares them with a wide range of data. Purpose is to obtain the best one for use in a water reactor safety computer simulator known as TRAC-B. Return to nucleate boiling can occur in a reactor accident at either high or low pressure and flow rates. Most of the correlations yield unrealistic results under some conditions. A new correlation is proposed which overcomes many of the deficiencies

  2. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  3. The results of feasibility study of co-generation NPP with innovative VK-300 simplified boiling water reactor

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.; Gabaraev, B. A.

    2004-01-01

    The co-generation nuclear power plant (CNPP) producing electricity and district-heating heat is planned to be constructed in Archangelsk Region of Russia. Following the Letter of Intent signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal /h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production.(author)

  4. A study on the jet pump characteristic curve in boiling water reactor

    International Nuclear Information System (INIS)

    Liao, L.Y.

    1990-01-01

    The jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From the investigation of the momentum equations, it is found that the normal quadrant jet pump models of these codes are essentially the same. In this paper, it is found that the relationship between the flow ratio, M, and the heat ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions the M - N characteristic curve will not change with power level, recirculation pump speed and loop flow rate. The effect of the gravitational head on the M - N curve has been found to be significant for low flow conditions. As a result, a guideline has been given to the definition of the specific energy (or the head ratio). Sensitivity studies on the key parameters have been performed. It is found that the generic M - N curve should not be used for a jet pump which does not have the same nozzle to throat area ratio as that of the generic jet pump

  5. Microgravity experiments on boiling and applications: research activity of advanced high heat flux cooling technology for electronic devices in Japan.

    Science.gov (United States)

    Suzuki, Koichi; Kawamura, Hiroshi

    2004-11-01

    Research and development on advanced high heat flux cooling technology for electronic devices has been carried out as the Project of Fundamental Technology Development for Energy Conservation, promoted by the New Energy and Industrial Technology Development Organization of Japan (NEDO). Based on the microgravity experiments on boiling heat transfer, the following useful results have obtained for the cooling of electronic devices. In subcooled flow boiling in a small channel, heat flux increases considerably more than the ordinary critical heat flux with microbubble emission in transition boiling, and dry out of the heating surface is disturbed. Successful enhancement of heat transfer is achieved by a capillary effect from grooved surface dual subchannels on the liquid supply. The critical heat flux increases 30-40 percent more than for ordinary subchannels. A self-wetting mechanism has been proposed, following investigation of bubble behavior in pool boiling of binary mixtures under microgravity. Ideas and a new concept have been proposed for the design of future cooling system in power electronics.

  6. Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-09-01

    single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.

  7. Fast critical experiment data for space reactors

    International Nuclear Information System (INIS)

    Collins, P.J.; McFarlane, H.F.; Olsen, D.N.; Atkinson, C.A.; Ross, J.R.

    1987-01-01

    Data from a number of previous critical experiments exist that are relevant to the design concepts being considered for SP-100 and MMW space reactors. Although substantial improvements in experiment techniques have since made some of the measured quantities somewhat suspect, the basic criticality data are still useful in most cases. However, the old experiments require recalculation with modern computational methods and nuclear cross section data before they can be applied to today's designs. Recently, we have calculated about 20 fast benchmark critical experiments with the latest ENDF/B data and modern transport codes. These calculations were undertaken as a part of the planning process for a new series of benchmark experiments aimed at supporting preliminary designs of SP-100 and MMW space reactors

  8. Signal processing techniques for sodium boiling noise detection

    International Nuclear Information System (INIS)

    1989-05-01

    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

  9. US graphite reactor D&D experience

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  10. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Volume 2. Appendices. Technical report, September 1977-October 1979

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWE. This volume contains the appendices.

  11. Water chemistry of nuclear reactor systems 5

    International Nuclear Information System (INIS)

    The water chemistry aspects of nuclear reactors are of critical importance. This book is a state-of-the-art review of the best international experience. It embodies the expertise presented at the fifth triennial international conference on the water chemistry of nuclear reactor systems in October 1989. The book is published in two volumes. Topics covered range widely and are grouped into eight sections. These include PWR experience (13 papers), radiation control measures (8 papers), BWR operational experience (10 papers), radiolysis in BWR coolants (11 papers), decontamination (10 papers), secondary-side chemistry (8 papers), water chemistry purification (8 papers), and fission product chemistry (4 papers). There are also 45 poster papers on aspects of water chemistry. All the papers are indexed separately. Discussion on the papers is included in volume 2 but is not indexed. (author)

  12. Liquid nitrogen - water interaction experiments for fusion reactor accident scenarios

    International Nuclear Information System (INIS)

    Duckworth, R.; Murphy, J.; Pfotenhauer, J.; Corradini, M.

    2001-01-01

    With the implementation of superconducting magnets in fusion reactors, the possibility exists for the interaction between water and cryogenic systems. The interaction between liquid nitrogen and water was investigated experimentally and numerically. The rate of pressurization and peak pressure were found to be driven thermodynamically by the expansion of the water and the boil-off of the liquid nitrogen and did not have a vapor explosion nature. Since the peak pressure was small in comparison to previous work with stratified geometries, the role of the geometry of the interacting fluids has been shown to be significant. Comparisons of the peak pressure and the rate of pressurization with respect to the ratio of the liquid nitrogen mass to water mass reveal no functional dependence as was observed in the liquid helium-water experiments. A simple thermodynamic model provides a fairly good description of the pressure rise data. From the data, the model will allow one to extract the interaction area of the water. As with previous liquid helium-water interaction experiments, more extensive investigation of the mass ratio and interaction geometry is needed to define boundaries between explosive and non-explosive conditions. (authors)

  13. Integrated plant safety assessment: Systematic Evaluation Program. LaCrosse Boiling Water Reactor, Dairyland Power Cooperative, Docket No. 50-409

    International Nuclear Information System (INIS)

    1983-04-01

    The Systematic Evaluation Program was initiated in February 1977 by the US Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides: (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. This report documents the review of the La Crosse Boiling Water Reactor, operated by Dairyland Power Cooperative. The La Crosse plant is one of 10 plants reviewed under Phase II of this program. This report indicates how 137 topics selected for review under Phase I of the program were addresed. Equipment and procedural changes have been identified as a result of the review

  14. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 2: A survey of the accuracy of the Studsvik of America CMS codes

    International Nuclear Information System (INIS)

    Demaziere, C.

    1999-02-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). The tools that are available to perform a modeling in the Department of Reactor Physics in Chalmers are CASMO-4/TABLES-3/SIMULATE-3 from Studsvik of America. Thus, before performing any kind of calculation with MOx fuels, it is necessary to be able to establish the reliability and the accuracy of these Core Management System (CMS) codes. This report presents a quantitative analysis of the models used in the package. A qualitative presentation is realized in a coming report

  15. The use of the average plutonium-content for criticality evaluation of boiling water reactor mixed oxide-fuel transport and storage packages

    International Nuclear Information System (INIS)

    Mattera, C.

    2003-01-01

    Currently in France, criticality studies in transport configurations for Boiling Water Reactor Mixed Oxide fuel assemblies are based on conservative hypothesis assuming that all rods (Mixed Oxide (Uranium and Plutonium), Uranium Oxide, Uranium and (Gadolinium Oxide rods) are Mixed Oxide rods with the same Plutonium-content, corresponding to the maximum value. In that way, the real heterogeneous mapping of the assembly is masked and covered by an homogenous Plutonium-content assembly, enriched at the maximum value. As this calculation hypothesis is extremely conservative, Cogema Logistics (formerly Transnucleaire) has studied a new calculation method based on the use of the average Plutonium-content in the criticality studies. The use of the average Plutonium-content instead of the real Plutonium-content profiles provides a highest reactivity value that makes it globally conservative. This method can be applied for all Boiling Water Reactor Mixed Oxide complete fuel assemblies of type 8 x 8, 9 x 9 and 10 x 10 which Plutonium-content in mass weight does not exceed 15%; it provides advantages which are discussed in the paper. (author)

  16. Westinghouse PWR and BWR reactor vessel segmentation experience in using mechanical cutting process

    International Nuclear Information System (INIS)

    Segerud, Per; Fallstroem, Stefan; Boucau, Joseph; Kreitman, Paul J.

    2011-01-01

    Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR's) and Boiling Water Reactors (BWR's) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date. (author)

  17. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hudong

    2002-12-09

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation.

  18. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Chen, Hudong

    2002-01-01

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation

  19. Reactors Project Delivery: The Value of Experiance

    International Nuclear Information System (INIS)

    Stosic, V. Zoran

    2014-01-01

    State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

  20. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  1. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    International Nuclear Information System (INIS)

    Venker, Jeanne

    2015-01-01

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO 2 ) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO 2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO 2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO 2 Brayton cycles and to evaluate the introduced heat removal system

  2. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    Science.gov (United States)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  3. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

  4. Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

    2017-06-15

    The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

  5. Towards a CFD model for boiling flows: validation of QMOM predictions with TOPFLOW experiments

    OpenAIRE

    Buffo, Antonio; Vanni, Marco; Marchisio, Daniele L.; Montoya, Gustavo; Baglietto, Emilio

    2017-01-01

    Boiling flows are very complex systems, usually confined in vertical pipes, where the liquid water moving upwards and the steam gas bubbles generated at the walls. The fluid dynamics of such systems is determined by the interplay of many different phenomena, including bubble nucleation, growth, condensation, coalescence, and breakage. For this reason, the development of a fully predictive computational fluid dynamics (CFD) model is very challenging, therefore we focus here only on some of the...

  6. Experience with valves for PHWR reactors

    International Nuclear Information System (INIS)

    Narayan, K.; Mhetre, S.G.

    1977-01-01

    Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

  7. Reactor G1: high power experiments

    International Nuclear Information System (INIS)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

    1957-01-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  8. Advanced test reactor testing experience-past, present and future

    International Nuclear Information System (INIS)

    Marshall, Frances M.

    2006-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

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

    International Nuclear Information System (INIS)

    Michiyoshi, I.; Takenaka, N.; Takahashi, O.

    1986-01-01

    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)

  10. Error analysis of the hydrodynamic vertical load function, 1/5-scale MKI BWR pressure suppression experiment and analysis program. Progress report

    International Nuclear Information System (INIS)

    McCauley, E.; Lai, W.; Carr, E.; Holman, G.

    1978-11-01

    Because of time constraints, the final report for the 1/5-scale Mark I boiling water reactor pressure suppression experiment presented results without critical analysis of the data. In particular, error bounds were not determined for the hydrodynamic vertical load function (HVLF). The objective of this study is to calculate the HVLF and the maximum download and maximum upload ratios for the 90 0 sector to 7.5 0 sector (i.e., the 3D to 2D ratios) together with their associated error bounds. Methodologies used to compute the HVLF, the 3D to 2D ratios, and error propagation for Test 1.3.1 are presented

  11. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  12. Application of Sub-cooled Boiling Model to Thermal-hydraulic Analysis Inside a CANDU-6 Fuel Channel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kang, Hyoung Chul; Yoo, Seong Yeon [Chungnam National Univ., Daejeon (Korea, Republic of)

    2007-07-01

    Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a sub-cooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. The objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code.

  13. Boiling curve in high quality flow boiling

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Hein, R.A.; Yadigaroglu, G.

    1980-01-01

    The post dry-out heat transfer regime of the flow boiling curve was investigated experimentally for high pressure water at high qualities. The test section was a short round tube located downstream of a hot patch created by a temperature controlled segment of tubing. Results from the experiment showed that the distance from the dryout point has a significant effect on the downstream temperatures and there was no unique boiling curve. The heat transfer coefficients measured sufficiently downstream of the dryout point could be correlated using the Heineman correlation for superheated steam, indicating that the droplet deposition effects could be neglected in this region

  14. CFD simulation of subcooled flow boiling at low pressure

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2001-01-01

    An increased interest to numerically simulate the subcooled flow boiling at low pressures (1 to 10 bar) has been aroused in recent years, pursued by the need to perform safety analyses of research nuclear reactors and to investigate the sump cooling concept for future light water reactors. In this paper the subcooled flow boiling has been simulated with a multidimensional two-fluid model used in a CFX-4.3 computational fluid dynamics (CFD) code. The existing model was adequately modified for low pressure conditions. It was shown that interfacial forces, which are usually used for adiabatic flows, need to be modeled to simulate subcooled boiling at low pressure conditions. Simulation results are compared against published experimental data [1] and agree well with experiments.(author)

  15. Experience with mechanical segmentation of reactor internals

    International Nuclear Information System (INIS)

    Carlson, R.; Hedin, G.

    2003-01-01

    Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

  16. Universality of oscillating boiling in Leidenfrost transition.

    Science.gov (United States)

    Khavari, Mohammad; Tran, Tuan

    2017-10-01

    The Leidenfrost transition leads a boiling system to the boiling crisis, a state in which the liquid loses contact with the heated surface due to excessive vapor generation. Here, using experiments of liquid droplets boiling on a heated surface, we report a phenomenon, termed oscillating boiling, at the Leidenfrost transition. We show that oscillating boiling results from the competition between two effects: separation of liquid from the heated surface due to localized boiling and rewetting. We argue theoretically that the Leidenfrost transition can be predicted based on its link with the oscillating boiling phenomenon and verify the prediction experimentally for various liquids.

  17. Enhanced heat transfer in confined pool boiling

    NARCIS (Netherlands)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    1982-01-01

    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

  19. Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.

    2007-01-01

    Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

  20. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  1. Customer-operator partnership. A boiling water reactor developed jointly by AREVA NP and E.ON Kernkraft

    International Nuclear Information System (INIS)

    Pasler, Doris; Gauthier, Jean Claude; Diercks, Frank; Fuchs, Michael

    2009-01-01

    Many countries spread all over the world have publicly expressed their intention to pursue the construction of new nuclear power plants with improved safety, economy and more straight forwarded operation and maintenance. Reasons for the intention are: The world wide increasing demand for energy and hence the general necessity to build new power plants. The concerns for increased emissions of green house gases leading to a change in the climate have brought into question the primary reliance on plants utilizing fossil fuels. A new reactor type matching the previously stated issues is AREVA NP's further development of proven BWR design. Combining AREVA's and E.ON's expertise, a project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1,250 MW, a net efficiency of about 37% and a design service life of 60 years. Within this joint venture the overall plant design was simplified and additionally all active safety systems have passive safety related backup systems utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. The development takes into account the technical and accumulated operating experience of the project partners. Based on the operating experience of the project partners a simplification of the overall system engineering was performed, flexible fuel cycle length (12 to 24 months) are possible as well as a reduction of process waste was achieved. These improvements regarding the operation and economics result on the one hand in lower investment cost and on the other hand in a high availability of the plant, hence in low maintenance costs. Generally, the electrical generation costs are accomplished, which are competitive to larger-capacity nuclear power plants and fossil-fired plants. (author)

  2. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de despresurizacion automatica (ADS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, C.; Chavez M, C., E-mail: ces.raga@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  3. Critical heat flux experiments for high conversion light water reactor, (3)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

    1990-03-01

    As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

  4. Start-up analysis of INET-5 MW district heating prototype reactor

    International Nuclear Information System (INIS)

    Li Tianshu

    1991-09-01

    The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

  5. Storage experience in Hungary with fuel from research reactors

    International Nuclear Information System (INIS)

    Gado, J.; Hargitai, T.

    1996-01-01

    In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

  6. Burnout experiment in subcooled forced-convection boiling of water for beam dumps of a high power neutral beam injector

    International Nuclear Information System (INIS)

    Horiike, Hiroshi; Kuriyama, Masaaki; Morita, Hiroaki

    1982-01-01

    Experimental studies were made on burnout heat flux in highly subcooled forced-convection boiling of water for the design of beam dumps of a high power neutral beam injector for Japan Atomic Energy Research Institute Tokamak-60. These dumps are composed of many circular tubes with two longitudinal fins. The tube was irradiated with nonuniformly distributed hydrogen ion beams of 120 to 200 kW for as long as 10 s. The coolant water was circulated at flow velocities of 3 to 7.5 m/s at exit pressures of 0.4 to 0.9 MPa. The burnout and film-boiling data were obtained at local heat fluxes of 8 to 15 MW/m 2 . These values were as high as 2.5 times larger than those for the circumferentially uniform heat flux case with the same parameters. These data showed insensitivity to local subcooling as well as to pressure, and simple burnout correlations were derived. From these results, the beam dumps have been designed to receive energetic beam fluxes of as high as 5 MW/m 2 with a margin of a factor of 2 for burnout

  7. Nuclear safety experiences of the research reactors at Boris Kidric Institute

    International Nuclear Information System (INIS)

    Pesic, M.; Cupac, S.; Stefanovic, D.

    1988-01-01

    Although the nuclear power was introduced recently in Yugoslavia with NPP KRSKO (connected to the grid in 1982.), nearly thirty years of wide experiences in nuclear safety is acquired at Boris Kidric Institute during operation of research reactors RA and RB. Reactor RB is an unshielded, zero power reactor, constructed in 1958, in as an actually bare critical natural uranium - heavy water system [1]. It was the first nuclear reactor designed in Yugoslavia and based on the national plan of development of nuclear energy [1,2,3,14]. Reactor RA is a 6.5 Mw heavy water research reactor designed in 1956, in USSR. Construction of the Yugoslav RA reactor started in 1957, and the reactor began operation in 1959, as one of the research reactors with better characteristics in the world at that time [4,5,6,14]. The main improvements in general and particularly nuclear safety both of the reactors in the past 30 years are given in this paper. These progressions are results of long experience acquired in reactor operation, maintenance, testing, development and application of newer safety criteria. The newer safety criteria have demanded improvements in the technical characteristics and in the administrative regulation domains both of the reactors

  8. Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

    Science.gov (United States)

    Gaikwad, Avinash J.; Vijayan, P. K.; Iyer, Kannan; Bhartiya, Sharad; Kumar, Rajesh; Lele, H. G.; Ghosh, A. K.; Kushwaha, H. S.; Sinha, R. K.

    2009-12-01

    For AHWR (Advanced Heavy Water Reactor), a pressure tube type Boiling Water Reactor (BWR) with parallel inter-connected loops, the Steam Drum (SD) level control is closely related to Main Heat Transport (MHT) coolant inventory and sustained heat removal through natural circulation, hence overall safety of the power plant. The MHT configuration with multiple (four) interconnected loops influences the SD level control in a manner which has not been previously addressed. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Event (PIEs) for Loss of Coolant Accident (LOCA), which postulates a double ended break in the four partitioned Emergency Core Cooling System (ECCS) header. A conventional individual three-element SD level controller can not account for the highly coupled and interacting behaviors, of the four SD levels. An innovative three-element SD level control scheme is proposed to overcome this situation. The response obtained for a variety of unsymmetrical disturbances shows that the SD levels do not diverge and quickly settle to the various new set points assigned. The proposed scheme also leads to enhanced safety margins for most of the PIEs considered with a little influence on the 100% full power steady-state design conditions.

  9. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  10. Technology, safety and costs of decommissioning a refernce boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation

  11. Nucleate boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    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)

  12. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

  13. Preliminary results of sodium boiling through a 19 heating rod bundle

    International Nuclear Information System (INIS)

    Menant, B.

    1975-01-01

    A test section including the GR.19 heating pin bundle has been designed in order to simulate a fast reactor sub-assembly. A first series of boiling experiments was performed with this text section on the CFNa II loop of the Service des Transferts Thermiques. Differences of temperature in the hottest section of the bundle were such that boiling was detected whereas the mean outlet temperature was more than 100 deg C below saturation. A study of the different aspects of undersaturated boiling was performed [fr

  14. Nonlinear stability models and analyses of the nuclear-coupled thermal-hydraulic behavior of boiling water reactors. Interim report, October 1, 1992--May 30, 1993

    International Nuclear Information System (INIS)

    Dorning, J.J.

    1993-05-01

    All the objectives originally scheduled for the first year of this grant have been achieved. Furthermore, the project is ahead of schedule, in that a substantial amount of work has been completed on two significant objectives originally planned for the second year. This interim report is divided into five parts, summarizing the mathematical development, analysis and results of the project goals -- goals originally planned for the first year and completed, and those on which substantial progress has been made ahead of schedule. Effects of unheated riser sections and the downcomer recirculation loop on the stability characteristics of advanced boiling water reactor designs that incorporate risers or unheated channel extensions are summarized in Part A. Such extensions are incorporated above the heated reactor core channels to enhance buoyancy-driven natural thermal convection both during normal at-power operation and during emergency shutdown. The effects of both, the inclusion of unheated riser sections in the designs (one of the goals substantially completed ahead of schedule), and the inclusion of the recirculation loop in the models (first year goal) were generally found to be destabilizing. In general, as riser lengths were increased equilibria that previously were stable became unstable, and the systems with the taller risers evolved to density-wave limit cycle oscillations. As a building block of the second year goal -- to extend the one dimensional dynamical analysis of reactor thermal-hydraulics/neutron-kinetics to two and three dimensions -- we have carried out, ahead of schedule, the nonlinear dynamical analysis of two-phase flow in multiple parallel heated channels. Some basic aspects of bifurcation phenomena in two-phase flow and the related nonlinear dynamics of single and multiple parallel, uniformly and nonuniformly heated channels are studied

  15. Experimental investigation on flow behavior during start-up of a heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Zhang Youjie

    1997-09-01

    An experimental simulation study on the transition from pressurized to boiling operation of a low-temperature, natural circulation nuclear heating reactor (5 MW) developed by INET of Tsinghua University is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and low-steam quality density wave instability on the transition from pressurized to boiling operation is described. The mechanism of flashing instability, which has never been studied well on this field, is especially interpreted. It is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: (1) increasing of initial pressure by means of a noncondensable gas (N 2 ), which is a very effective method to eliminate geyser instability and flashing instability at lower pressure. (2)start-up of the reactor at this pressurized condition with a constant heat flux under the limited value of q = 0.15 MW·m -2 , which controls the exit temperature of the heated section below the one of net vapor generation, the low steam quality density wave oscillation can be avoided. (3) transition to a lower pressure, boiling operation. The method of transition with low-heat flux and low-inlet subcooling is proposed: at pressurized operation condition, by reducing the heat flux to its lowest level, releasing the noncondensable gas and increasing the heat flux gradually (dq/dt -2 ·min -1 ), during which the low-steam quality density wave oscillation can be prevented from occurring, then the boiling operation condition can be achieved through adjusting the heat flux and inlet subcooling to their designed value. A stable transition from pressurized to boiling operation of the 5 MW reactor is achieved by careful selection of the thermohydraulic parameters. (7 refs., 7 figs., 1

  16. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  17. Utilization experience with research reactors of various power levels

    International Nuclear Information System (INIS)

    Chaudri, S.A.; Waheed, S.; Ahmad, S.

    1999-01-01

    Utilization of research reactor, PARR-1 at the power level of 5 MW, and then at 9 MW, after the up gradation and conversion from highly enriched uranium (HEU) to low enriched uranium (LEU), has been described. In addition, the type of work carried out around a smaller facility, PARR-2, with power rating of 27kW has also been discussed. Utilization of PARR-1 in the area of neutron diffraction, neutron capture studies, radioisotope production, neutron activation analysis, reactor physics, and in reactor controls etc has been illustrated. The benefits derived from the up gradation of the power for neutron diffraction studies, radioisotope production and neutron radiography have been discussed. The problem, which can be handed successfully on PARR-2, include neutron activation analysis, production of short-lived radioisotopes, and experimentation in reactor engineering and physics for training purposes. Suitable methodologies have been developed for the analysis of samples of varied nature using neutron activation technique, and the experience gained has been applied to the analysis of geological, environmental and biological samples and high purity materials. (author)

  18. Data acquisition. GRAAL experiment. Hybrid reactor experiment. AMS experiment

    International Nuclear Information System (INIS)

    Barancourt, D.; Barbier, G.; Bosson, G.; Bouvier, J.; Gallin-Martel, L.; Meillon, B.; Stassi, P.; Tournier, M.

    1997-01-01

    The main activity of the data acquisition team has consisted in hardware and software developments for the GRAAL experiment with the trigger board, for the 'Reacteurs Hybrides' group with an acquisition board ADCVME8V and for the AMS experiment with the monitoring of the aerogel detector. (authors)

  19. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  20. Categorization of core-damage sequences by containment event tree analysis for boiling water reactor with Mark-II containment

    International Nuclear Information System (INIS)

    Watanabe, N.; Kajimoto, M.; Muramatsu, K.

    1994-01-01

    In the present study, containment responses to core damage accidents were analyzed for a large spectrum of core damage sequences, which were defined by front-line system event trees, in a BWR with Mark-11 containment by using the Accident Progression Event Tree (APSET) method and their characteristics were examined in terms of mainly probabilistic aspects such as their respective conditional probabilities of containment failure modes and accident termination. This paper showed that various core damage sequences could be categorized into a small number of groups, each of which consisted of the sequences with similar containment response characteristics, as follows: Interfacing system LOCA; ATWS with high pressure injection available; Loss of long-term containment heat removal; Station blackout; Loss of coolant injection with the reactor not depressurized; Loss of coolant injection with the reactor depressurized; Loss of short-term containment heat removal; and Reactor pressure vessel rupture. The above categorization provides a perspective on the potential containment failure modes and the effectiveness of some accident mitigative measures, which could be useful for studying accident management strategies and as well for assisting the analysts in carrying out future CET analyses. (author)

  1. Subcooled boiling heat transfer in a short vertical SUS304-tube at liquid Reynolds number range 5.19 x 104 to 7.43 x 105

    International Nuclear Information System (INIS)

    Hata, Koichi; Masuzaki, Suguru

    2009-01-01

    The subcooled boiling heat transfer and the steady-state critical heat fluxes (CHFs) in a short vertical SUS304-tube for the flow velocities (u = 17.28-40.20 m/s), the inlet liquid temperatures (T in = 293.30-362.49 K), the inlet pressures (P in = 842.90-1467.93 kPa) and the exponentially increasing heat input (Q = Q 0 exp(t/τ), τ = 8.5 s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tubes of inner diameters (d = 3 and 6 mm), heated lengths (L = 33 and 59.5 mm), effective lengths (L eff = 23.3 and 49.1 mm), L/d (=11 and 9.92), L eff /d (=7.77 and 8.18), and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) are used in this work. The inner surface temperature and the heat flux from non-boiling to CHF are clarified. The subcooled boiling heat transfer for SUS304 test tube is compared with our Platinum test tube data and the values calculated by other workers' correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details and the widely and precisely predictable correlation of the subcooled boiling heat transfer for turbulent flow of water in a short vertical SUS304-tube is given based on the experimental data. The correlation can describe the subcooled boiling heat transfer obtained in this work within 15% difference. Nucleate boiling surface superheats for the SUS304 test tube become very high. Those at the high flow velocity are close to the lower limit of Heterogeneous Spontaneous Nucleation Temperature. The dominant mechanisms of the flow boiling CHF in a short vertical SUS304-tube are discussed.

  2. Reactor for exothermic reactions

    Science.gov (United States)

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

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  3. Simulation study of neutronic and acoustic methods of boiling detection

    International Nuclear Information System (INIS)

    Shimazaki, J.; Fujii, Y.; Shinohara, Y.; Araki, H.

    1981-01-01

    As a part of the work on the development of a diagnostic system for detecting anomalies in the core of a fast breeder reactor, a study is being made on the methods of detecting abnormal reactivity and coolant boiling on the basis computer simulation. The objective of the present simulation study is to obtain some basic information useful for developing sensitive method of signal processing suitable for boiling detection. Neutronic, acoustic and measurement noises as well as reactor kinetics and reactivity estimator are simulated on a hybrid computer and analyzed for various noise conditions using correlation, spectral and coherence analyses. It has been confirmed that the cross-spectral and coherence analyses between neutronic and acoustic signals are effective method for boiling detection provided that the void coefficient of reactivity is sufficiently large and that detection capability is slightly improved by using a discriminator of signal level before making spectral analysis. A simulated boiling detection experiment using a zero power thermal reactor is also planned in order to check the validity of computer simulation. (author)

  4. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  5. Proposed Reactor Operating Experience Feedback System Development

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Kim, Min Chul; Huh, Chang Wook; Lee, Durk Hun; Bae, Koo Hyun

    2006-01-01

    Most events occurring in nuclear power plants are not individually significant, and prevented from progressing to accident conditions by a series of barriers against core damage and radioactive releases. Significant events, if occur, are almost always a breach of these multiple barriers. As illustrated in the 'Swiss cheese' model, the individual layers of defense or 'cheese slices' have weakness or 'holes.' These weaknesses are inconstant, i.e., the holes are open or close at random. When by chance all the holes are aligned, a hazard causes the significant event of concern. Elements of low significant events, inattention to detail, time or economic pressure, uncorrected poor practices/habits, marginal maintenance and equipment care, etc., make holes in the layers of defense; some elements may make more holes in different layers, incurring more chances to be aligned. An effective reduction of the holes, therefore, is gained through better knowledge or awareness of increasing trends of the event elements, followed by appropriate actions. According to the Swiss cheese metaphor, attention to the Operating Experience (OE) feedback system, as opposed to the individual and to randomness, is drawn from a viewpoint of reactor safety

  6. Application of Subcooled Boiling Model to Thermal-hydraulic Analysis inside a CANDU-6 Fuel Channel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Yu, Jin Bok; Yi, Sung Chul [Hangyang Univ., Seoul (Korea, Republic of)

    2006-07-01

    Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a subcooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. Therefore, the objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code. The RPI model has been implemented in FLUENT 6.2 via user-defined functions (UDFs) in conjunction with the Eulerian multiphase model.

  7. Economic simplified boiling water reactor (ESBWR) response to an extended station blackout/ loss of all AC power

    International Nuclear Information System (INIS)

    Barrett, A.J.; Marquino, W.

    2013-01-01

    U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackout for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event

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

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, T.H.J.J. van der; Stekelenburg, A.J.C. [Delft Univ. of Technology (Netherlands)

    1995-09-01

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

  9. Boiling induced mixed convection in cooling loops

    International Nuclear Information System (INIS)

    Knebel, J.U.; Janssens-Maenhout, G.; Mueller, U.

    2000-01-01

    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)

  10. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

    1992-01-01

    Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

  11. E-chem page: A Support System for Remote Diagnosis of Water Quality in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Naohiro Kusumi; Takayasu Kasahara; Kazuhiko Akamine; Kenji Tada; Naoshi Usui; Nobuyuki Oota

    2002-01-01

    It is important to control and maintain water quality for nuclear power plants. Chemical engineers sample and monitor reactor water from various subsystems and analyze the chemical quality as routine operations. With regard to controlling water quality, new technologies have been developed and introduced to improve the water quality from both operation and material viewpoints. To maintain the quality, it is important to support chemical engineers in evaluating the water quality and realizing effective retrieval of stored data and documents. We have developed a remote support system using the Internet to diagnose BWR water quality, which we call e-chem page. The e-chem page integrates distributed data and information in a Web server, and makes it easy to evaluate the data on BWR water chemistry. This system is composed of four functions: data transmission, water quality evaluation, inquiry and history retrieval system, and reference to documents on BWR water chemistry. The developed system is now being evaluated in trial operations by Hitachi, Ltd. and an electric power company. In addition diagnosis technology applying independent component analysis (ICA) is being developed to improve predictive capability of the system. This paper describes the structure and function of the e-chem page and presents results of obtained with the proposed system for the prediction of chemistry conditions in reactor water. (authors)

  12. Fast breeder reactors: can we learn from experience

    International Nuclear Information System (INIS)

    Keck, O.

    1981-01-01

    An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

  13. Pool Boiling CHF in Inclined Narrow Annuli

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2010-01-01

    Pool boiling heat transfer has been studied extensively since it is frequently encountered in various heat transfer equipment. Recently, it has been widely investigated in nuclear power plants for application to the advanced light water reactors designs. Through the review on the published results it can be concluded that knowledge on the combined effects of the surface orientation and a confined space on pool boiling heat transfer is of great practical importance and also of great academic interest. Fujita et al. investigated pool boiling heat transfer, from boiling inception to the critical heat flux (CHF, q' CHF ), in a confined narrow space between heated and unheated parallel rectangular plates. They identified that both the confined space and the surface orientation changed heat transfer much. Kim and Suh changed the surface orientation angles of a downward heating rectangular channel having a narrow gap from the downward-facing position (180 .deg.) to the vertical position (90 .deg.). They observed that the CHF generally decreased as the inclination angle (θ ) increased. Yao and Chang studied pool boiling heat transfer in a confined heat transfer for vertical narrow annuli with closed bottoms. They observed that when the gap size ( s ) of the annulus was decreased the effect of space confinement to boiling heat transfer increased. The CHF was occurred at much lower value for the confined space comparing to the unconfined pool boiling. Pool boiling heat transfer in narrow horizontal annular crevices was studied by Hung and Yao. They concluded that the CHF decreased with decreasing gap size of the annuli and described the importance of the thin film evaporation to explain the lower CHF of narrow crevices. The effect of the inclination angle on the CHF on countercurrent boiling in an inclined uniformly heated tube with closed bottoms was also studied by Liu et al. They concluded that the CHF reduced with the inclination angle decrease. A study was carried out

  14. The reactor antineutrino anomaly and low energy threshold neutrino experiments

    Science.gov (United States)

    Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

    2018-01-01

    Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

  15. French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

    International Nuclear Information System (INIS)

    Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

    1996-01-01

    The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

  16. Brookhaven Reactor Experiment Control Facility, a distributed function computer network

    International Nuclear Information System (INIS)

    Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

    1975-11-01

    A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

  17. Experience in Reviewing Small Modular Reactor Technology

    International Nuclear Information System (INIS)

    Ahmad Nabil Abdul Rahim; Alfred, S.L.; Phongsakorn, P.

    2015-01-01

    Malaysia is in the stage of conducting Preliminary Technical Feasibility Study for the Deployment of Small Modular Reactor (SMR). There are different types of SMR, some already under construction in Argentina (CAREM) and China (HTR-PM) - (light water reactor and high temperature reactor technologies), others with near-term deployment such as SMART in South Korea, ACP100 in China, mPower and NuScale in the US, and others with longer term deployment prospects (liquid-metal cooled reactor technologies). The study was mainly to get an overview of the technology available in the market. The SMR ranking in the study was done through listing out the most deployable technology in the market according to their types. As a new comer country, the proven technology with an excellent operation history will usually be the main consideration points. (author)

  18. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    Hernandez M, J.L.; Ortiz V, J.

    2005-01-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  19. Once-through thorium fuel cycle evaluation for TVA's Browns Ferry-3 Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, G.C. (comp.)

    1982-05-01

    This report documents benchmark evaluations to test thorium lattice predictive methods and neutron cross sections against available data and summarizes specific evaluations of the once-through thorium cycle when applied to the Browns Ferry-3 BWR. It was concluded that appreciable uncertainties in thorium cycle nuclear data cloud the ability to reliably predict the fuel cycle performance and that power reactor irradiations of ThO/sub 2/ rods in BWRs are desirable to resolve uncertainties. Benchmark evaluations indicated that the ENDF/B-IV data used in the evaluations should cause an underprediction of U-233/ThO/sub 2/ fuel reactivity, and, therefore, the results of the preliminary evaluations completed under the program should be conservative.

  20. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  1. An Improved Design of a Simple Tubular Reactor Experiment.

    Science.gov (United States)

    Asfour, Abdul-Fattah A.

    1985-01-01

    Background information, procedures used, and typical results obtained are provided for an experiment which: (1) examines the effect of residence time on conversion in a tubular flow reactor; and (2) compares the experimental conversions with those obtained from plug-flow and laminar-flow reactor models. (JN)

  2. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  3. Changes in phytochemicals, anti-nutrients and antioxidant activity in leafy vegetables by microwave boiling with normal and 5% NaCl solution.

    Science.gov (United States)

    Singh, Shrawan; Swain, S; Singh, D R; Salim, K M; Nayak, Dipak; Roy, S Dam

    2015-06-01

    The present study investigated the changes in phytochemicals and antioxidant activities in 25 leafy vegetables with two common boiling practices viz., with 5% NaCl solution (BSW) and normal water (BNW) in a domestic microwave oven. Fresh samples (100g) were rich in polyphenols (58.8-296.9mg), tannin (402.0-519.4mg), flavonoids (148.9-614.4mg), carotenoids (69.0-786.3mg), anthocyanin (11.7-493.7mg) and ascorbic acid (245.0-314.2mg). Microwave boiling significantly (pvegetables. Boiling process reduced anti-nutrients from fresh samples (FS) as observed for nitrate (4.5-73.6% by BSW and 22.5-98.8% by BNW); phytate (6.2-69.7% by BSW and 10.6-57.3% by BNW) and oxalate (14.7-88.9% by BSW and 14.5-87.3% by BNW) but saponin increased in 18 vegetables by BNW while 8 vegetables by BSW. The study revealed differential pattern of change in phytochemical matrix and anti-nutrients in vegetables by microwave boiling which will help in devising efficient cooking practices and contribute in health and nutritional security. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. New strategies of reloads design and models of control bars in boiling water reactors; Nuevas estrategias de diseno de recargas y de patrones de barras de control en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    In this work the results obtained when analyzing new strategies in the reload designs of nuclear fuel and models of control bars, for boiling water reactors are presented. The idea is to analyze the behaviour of the reactor during an operation cycle, when the heuristic rules are not used (commonly used by expert engineers in both designs). Specifically was analyzed the rule of low leak and the load strategy Control Cell Core for the design of a fuel reload. In a same way was analyzed the rule of prohibiting the use of the intermediate positions in the control bars, as well as the construction of bar models based on load strategies type Control Cell Core. In the first analysis a balance and transition cycle were used. For the second analysis only a transition cycle was used, firstly with the reloads designed in the first analysis and later on with reloads built by other methods. For the simulation of the different configurations proposed in both cases, was used the code Simulate-3. To obtain the designs in both studies, the heuristic techniques or neural networks and taboo search were used. The obtained results show that it can be omitted of some rules used in the ambit for the mentioned designs and even so to obtain good results. To carry out this investigation was used Dell work station under Li nux platform. (Author)

  5. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  6. Use of an influence diagram and fuzzy probability for evaluating accident management in a boiling water reactor

    International Nuclear Information System (INIS)

    Yu, D.; Kastenberg, W.E.; Okrent, D.

    1994-01-01

    A new approach is presented for evaluating the uncertainties inherent in severe accident management strategies. At first, this analysis considers accident management as a decision problem (i.e., applying a strategy compared with do nothing) and uses an influence diagram. To evaluate imprecise node probabilities in the influence diagram, the analysis introduces the concept of a fuzzy probability. When fuzzy logic is applied, fuzzy probabilities are easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach, which uses point-estimate values, but also additional information regarding the impact of using imprecise input data. As an illustrative example, the proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence at the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy is beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of containment failure for both liner melt-through and late overpressurization. Even though uncertainty exists in the results, flooding is preferred to do nothing when evaluated in terms of two risk measures: early and late fatalities

  7. On the Modeling of Local Neutronically-Coupled Flow-Induced Oscillations in Advanced Boiling Water Reactors

    International Nuclear Information System (INIS)

    Aniel-Buchheit, Sylvie; Podowski, Michael Z.

    2006-01-01

    The purpose of this paper is to discuss the development in progress of a complete space- and time-dependent model of the coupled neutron kinetic and reactor thermal-hydraulics. The neutron kinetics model is based on two-group diffusion equations with Doppler and void reactivity feedback effects. This model is coupled with the model of two-phase flow and heat transfer in parallel coolant channels. The modeling concepts considered for this purpose include one-dimensional drift flux and two-fluid models, as well a CFD model implemented in the NPHASE advanced computational multiphase fluid dynamics (CMFD) computer code. Two methods of solution for the overall model are proposed. One is based on direct numerical integration of the spatially-discretized governing equations. The other approach is based on a quasi-analytical modal approach to the neutronics model, in which a complete set of eigenvectors is found for step-wise temporal changes of the cross-sections of core materials (fuel and coolant/moderator). The issues investigated in the paper include details of model formulation, as well as the results of calculations for neutronically-coupled density-wave oscillations. (authors)

  8. PSI-BOIL, a building block towards the multi-scale modeling of flow boiling phenomena

    International Nuclear Information System (INIS)

    Niceno, Bojan; Andreani, Michele; Prasser, Horst-Michael

    2008-01-01

    Full text of publication follows: In these work we report the current status of the Swiss project Multi-scale Modeling Analysis (MSMA), jointly financed by PSI and Swissnuclear. The project aims at addressing the multi-scale (down to nano-scale) modelling of convective boiling phenomena, and the development of physically-based closure laws for the physical scales appropriate to the problem considered, to be used within Computational Fluid Dynamics (CFD) codes. The final goal is to construct a new computational tool, called Parallel Simulator of Boiling phenomena (PSI-BOIL) for the direct simulation of processes all the way down to the small-scales of interest and an improved CFD code for the mechanistic prediction of two-phase flow and heat transfer in the fuel rod bundle of a nuclear reactor. An improved understanding of the physics of boiling will be gained from the theoretical work as well as from novel small- and medium scale experiments targeted to assist the development of closure laws. PSI-BOIL is a computer program designed for efficient simulation of turbulent fluid flow and heat transfer phenomena in simple geometries. Turbulence is simulated directly (DNS) and its efficiency plays a vital role in a successful simulation. Having high performance as one of the main prerequisites, PSIBOIL is tailored in such a way to be as efficient a tool as possible, relying on well-established numerical techniques and sacrificing all the features which are not essential for the success of this project and which might slow down the solution procedure. The governing equations are discretized in space with orthogonal staggered finite volume method. Time discretization is performed with projection method, the most obvious a the most widely used choice for DNS. Systems of linearized equation, stemming from the discretization of governing equations, are solved with the Additive Correction Multigrid (ACM). methods. Two distinguished features of PSI-BOIL are the possibility to

  9. System Requirements Document for the Molten Salt Reactor Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Aigner, R.D.

    2000-04-01

    The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

  10. Status of the INL gas reactor test system experiment facility

    International Nuclear Information System (INIS)

    Marshall, Theron; Bennet, Brion; Tschaggeny, Charles; Reyes, Jose; Groome, John

    2007-01-01

    The Gas Reactor Test System (GRTS) is an experiment facility for examining the thermal hydraulic performance of the Generation IV, Very High Temperature Reactor (VHTR) during a Large-Break Loss of Coolant Accident (LB-LOCA). The LB-LOCA is defined as the double guillotine break of the VHTR coaxial inlet and outlet cross duct. Two system safety codes, MELCOR and RELAP5-3D were used to calculate core temperatures and flow rates during the LB-LOCA transient. Computational fluid dynamics modeling of the transient produced flow vectors and gas species distribution. The most important phenomenon during the transient is the lock-exchange process, which suppresses the onset of natural circulation until considerable molecular diffusion has occurred. The GRTS was designed based upon a hierarchical two tier scaling analysis whose primary objective was replicating the lock-exchange and natural circulation characteristics of the VHTR. The GRTS uses a scaled graphite core to represent the VHTR's graphite core. An in-depth scaling analysis was performed for the GRTS in order to ensure that it accurately simulated the VHTR thermal responses. RELAP5-3D thermal analyses, ProEngineer stress analyses, and combined FLUENT-STARCD CFD analyses have provided a system design that fulfills the GRTS mission statement. This paper discusses the design analyses and their implications on the GRTS capabilities. A discussion is also presented on the preliminary instrumentation plan. The GRTS will provide an extensive temperature map of the VHTR core outlet plenum and its core support, oxygen transport rates during the lock-exchange phenomenon, and thermal conduction rates from the core to the vessel. As a result of the GRTS using helium coolant at 950 C, the resulting experiment data is expected to considerably extend the U.S. database for high-temperature gas reactor operations. Finally, the discussion will present conclusions from the GRTS manufacturing and quality control processes that may

  11. Comparison of Computational Results with a Low-g, Nitrogen Slosh and Boiling Experiment

    Science.gov (United States)

    Stewart, Mark; Moder, Jeff

    2015-01-01

    The proposed paper will compare a fluid/thermal simulation, in FLUENT, with a low-g, nitrogen slosh experiment. The French Space Agency, CNES, performed cryogenic nitrogen experiments in several zero gravity aircraft campaigns. The computational results have been compared with high-speed photographic data, pressure data, and temperature data from sensors on the axis of the cylindrically shaped tank. The comparison between these experimental and computational results is generally favorable: the initial temperature stratification is in good agreement, and the two-phase fluid motion is qualitatively captured.

  12. Radiation induced boiling enhancement. 5th report, Leidenfrost temperature and quenching condition

    Energy Technology Data Exchange (ETDEWEB)

    Imai, Yasuyuki; Okamoto, Koji [Tokyo Univ., Tokai, Ibaraki (Japan); Fukamachi, Norihiro; Takamasa, Tomoji [Tokyo Univ. of Mercantile Marine, Tokyo (Japan); Mishima, Kaichiro [Kyoto Univ., Kumatori, Osaka (Japan); Furuya, Masahiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2002-07-01

    Improvement of quenching requires that the cooling liquid can contact the heating surface, or a high-wettability heating surface, even if a vapor film is generated on the surface. From this point of view, an experimental study to investigate Leidenfrost condition and quenching was performed by use of an oxide semiconductor-coated material under a {gamma} ray irradiation environment. The results showed that Leidenfrost condition of oxidized titanium was improved up to 30-50degC after 800kGy {sup 60}Co {gamma} ray irradiated. Liquid film velocity flowing down a heated vertical wall was estimated using the result of the experiment. (author)

  13. DOE's foreign research reactor transportation services contract: Perspective and experience

    International Nuclear Information System (INIS)

    Patterson, John

    1997-01-01

    DOE committed to low- and moderate-income countries participating in the foreign research reactor spent fuel returns program that the United States government would provide for the transportation of the spent fuel. In fulfillment of that commitment, DOE entered into transportation services contracts with qualified, private-sector firms. NAC will discuss its experience as a transportation services provider, including range of services available to the foreign reactors, advantages to DOE and to the foreign research reactors, access to contract services by high income countries and potential advantages, and experience with initial tasks performed under the contract. (author)

  14. Analysis of the peach bottom 2 BWR turbine trip experiment by RELAP 5/3.2 code

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2002-01-01

    Full Text Available This paper presents the results of the application of the system of the thermalhydraulic code RELAP5/Mod3.2 in predicting the Peach Bottom Boiling Water Reactor Turbine Trip test. This experiment constitutes a challenge to the capabilities of current computational tools in realistically predicting transient scenarios in nuclear power plants. In fact, it involves strong feedback during the transient between thermalhydraulics and neutronics. In this respect, a reference case was run in order to simulate the interactions between the generated steam line pressure wave propagation and the instantaneous core void distribution. An overall comparison shows good agreement between the code calculations and the experimental data. A series of sensitivity analyses were also performed in order to assess the code prediction features, as well as to identify uncertainties related to the adopted thermalhydraulic parameters used for the plant modelisation.

  15. Ageing management experience at NUR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melllal, Sabrina; Rezig, Mohamed; Zamoun, Rachid; Ameur, Azeddin [Nuclear Research Center of Draria, Algiers (Algeria)

    2013-07-01

    NUR is a 1 MW, open pool reactor moderated and cooled by light water. It was commissioned in 1989. NUR is used for education and training in Nuclear Engineering and related topics for COMENA and National Scientific Community. It is also used to perform R and D works and services at national and regional levels. In this presentation, we describe the methodology and the main development activities related to the ageing management at NUR reactor. These activities include inspection actions and development actions to introduce modifications, to solve obsolescence issues in view to implement the required preventive and curative maintenance programs and to improve the performances of the installation. These actions involved mainly the Operation Assistance System of the Reactor (OAS), the secondary cooling loop, the cooling tower. A new OAS using a new technology and having more possibilities than the older one was introduced in the control system of the reactor. The OAS hardware structure, software structure and the main functions performed are presented. The second loop is entirely refurbished. Two new cooling towers are installed and connected to the main heat exchanger with new piping and valves. The architecture of this new installation is described and the performance assessed. Other actions which involve auxiliary systems like emergency electrical system, air pneumatic system and automatic fire extinguishing are presented.

  16. Physical Science Informatics: Providing Open Science Access to Microheater Array Boiling Experiment Data

    Science.gov (United States)

    McQuillen, John; Green, Robert D.; Henrie, Ben; Miller, Teresa; Chiaramonte, Fran

    2014-01-01

    The Physical Science Informatics (PSI) system is the next step in this an effort to make NASA sponsored flight data available to the scientific and engineering community, along with the general public. The experimental data, from six overall disciplines, Combustion Science, Fluid Physics, Complex Fluids, Fundamental Physics, and Materials Science, will present some unique challenges. Besides data in textual or numerical format, large portions of both the raw and analyzed data for many of these experiments are digital images and video, requiring large data storage requirements. In addition, the accessible data will include experiment design and engineering data (including applicable drawings), any analytical or numerical models, publications, reports, and patents, and any commercial products developed as a result of the research. This objective of paper includes the following: Present the preliminary layout (Figure 2) of MABE data within the PSI database. Obtain feedback on the layout. Present the procedure to obtain access to this database.

  17. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure - main report. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J.; McDuffie, P.N.

    1996-07-01

    The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s Washington Nuclear Plant Two (WNP-2), which is a boiling water reactor (BWR), located at Richland, Washington, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a {open_quotes}green field{close_quotes} condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low- level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities. Sensitivity of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances is also examined.

  18. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

    1982-01-01

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  20. Analysis of the rotation accident of assemblies in boiling water reactors; Analisis del accidente de rotacion de ensambles en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Becerril-Gonzalez M, J. J. [Universidad Autonoma de Yucatan, Av. Industrias no contaminantes por Anillo Periferico Norte s/n, Apdo. Postal 150 Cordemex, Merida, Yucatan (Mexico); Fuentes M, L.; Castillo M, J. A.; Ortiz S, J. J.; Perusquia de Cueto, R., E-mail: juanjosebecerril_1@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    For this work was analyzed the impact that would cause the load of a rotated fuel assembly in the behaviour of the core in the Cycle 14 of the Unit 1 of the nuclear power plant of Laguna Verde. To carry out this analysis the code Simulate-3 was used, with which was possible to analyze the behavior of the effective multiplication factor and the thermal limits (MAPRAT, MFLPD and MFLCPR). The rotation of fuel assemblies to 90, 180 and 270 grades was analyzed with regard to the design position, with 0, 1, 2 and 3 burnt cycles for these assemblies. The results show that the thermal limits remain inside the allowed values, therefore if this accident type happened the reactor could continue operating in a sure way. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-15

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

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

    International Nuclear Information System (INIS)

    1994-01-01

    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

  3. Qualification of the Darwin code for the studies of the fuel cycle relative to the boiling water reactors; Qualification du formulaire Darwin pour les etudes du cycle du combustible pour les reacteurs a eau bouillante

    Energy Technology Data Exchange (ETDEWEB)

    Allais, V

    1998-03-01

    This thesis was carried out in the framework of fuel cycles studies in partnership with COGEMA; the aim is to determine physics parameters characterising Boiling Reactor Assemblies. Those reactors Firstly distinguish themselves from Pressurised Water Reactor by the boiling of the moderator in the core and secondary by the strong neutronics heterogeneity due to complex design. The diphasic mixture formed is characterised by the void fraction parameter. The loss of information, and neutronic studies characteristics of Boiling Water Reactors led us to make preliminary studies having in view to quantify the void fraction impact on the isotopics evolution. Studies on neutronics influence of assemblies and control rods from the immediate environment allows to define the cluster size to describe. The radial description optimisation with APOLLO-2 is necessary to improve the calculation performance and to reduce the errors coming from the modelization. The following points were studied: pellet radial discretization, clustering of cells characterized by a similar behaviour, options in flux spatial calculation (interface current formalism), self-shielding optimisation (specific to each isotopes). The three dimensional modelization with CRONOS-2 and the simplified accounting of the thermohydraulics / neutronics coupling done by a procedure developed and written during this thesis, allow an evaluation of axial distribution of void fraction, power and burn-up during the irradiation. The comparison with experimental analytic results of complete assembly and pin samples dissolutions allows the qualification of this procedure and confirms the necessity to take into account the void fraction axial variation during the evolution. The application of an automatic coupling with the DARWIN cycle code will allow a precise burnup calculation to be utilized in an industrial procedure. (author)

  4. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  5. Expanding of reactor power calculation model of RELAP5 code

    International Nuclear Information System (INIS)

    Lin Meng; Yang Yanhua; Chen Yuqing; Zhang Hong; Liu Dingming

    2007-01-01

    For better analyzing of the nuclear power transient in rod-controlled reactor core by RELAP5 code, a nuclear reactor thermal-hydraulic best-estimate system code, it is expected to get the nuclear power using not only the point neutron kinetics model but also one-dimension neutron kinetics model. Thus an existing one-dimension nuclear reactor physics code was modified, to couple its neutron kinetics model with the RELAP5 thermal-hydraulic model. The detailed example test proves that the coupling is valid and correct. (authors)

  6. Experiments during flow boiling of a R22 drop-in: R422D adiabatic pressure gradients

    International Nuclear Information System (INIS)

    Rosato, A.; Mauro, A.W.; Mastrullo, R.; Vanoli, G.P.

    2009-01-01

    R22, the HCFC most widely used in refrigeration and air-conditioning systems in the last years, is phasing-out. R422D, a zero ozone-depleting mixture of R125, R134a and R600a (65.1%/31.5%/3.4% by weight, respectively), has been recently proposed as a drop-in substitute. For energy consumption calculations and temperature control, it is of primary importance to estimate operating conditions after substitution. To determine pressure drop in the evaporator and piping line to the compressor, in this paper the experimental adiabatic pressure gradients during flow boiling of R422D are reported for a circular smooth horizontal tube (3.00 mm inner radius) in a range of operating conditions of interest for dry-expansion evaporators. The data are used to establish the best predictive method for calculations and its accuracy: the Moreno-Quiben and Thome method provided the best predictions for the whole database and also for the segregated data in the annular flow regime. Finally, the experimental data have been compared with the adiabatic pressure gradients of both R22 and its much used alternative R407C available in the literature.

  7. Crack initiation in the Nb-stabilized austenitic steel (A347) in the core shroud and top and core guide of a german boiling water reactor - description of the extent of the damage and explanation of its causes

    International Nuclear Information System (INIS)

    Wachter, O.; Bruns, J.; Wesseling, U.; KIlian, R.; Roth, A.

    1998-01-01

    Depending on the material state, stabilized austenitic steels can be susceptible to intergranular stress corrosion cracking (IGSCC) under the operating conditions of a boiling water reactor (BWR). This surprising experience for German reactor technology over the last three years arose from the observation of cracks, Firstly in the hot water piping of Ti-stabilized austenitic steel A321 in six BWR plants and later in the reactor pressure vessel internals of Nb-stabilized austenitic steel A347 in one BWR plant. In this report, the findings concerning the core shroud (upper and lower support rings) and the top and core guide itself are described. The results of the visual inspection, ultrasonic testing and the microstructure are presented and discussed with respect to the cause of the damage. In all cases, the damage in the core shroud and the top and core guides was ascribed to IGSCC, following chromium depletion at the grain boundaries (sensitization). This Sensitization was caused by a stress relief heat treatment of the support rings of the core shroud and the reinforcing rings of the top and core guide, all of which were made from the same heat. This heat exhibited a high free carbon content (high carbon content with low degree of stabilization by Nb) which led to the precipitation of Cr 23 C 6 at the grain boundaries during heat treatment. Residual welding stresses provided the tensile stresses necessary of IGSCC - the service stresses on the components were low and were considered to be only of minor importance. With regard to the corrosive medium, in addition to the conductivity, the influence of the corrosion potential which was mainly determined by the radiolytic formation of H 2 O 2 was recognized. As solution to the problem, the application of steels of low carbon content with the maximum allowable stabilization ration and optimized production processes (heat input to be as low as possible or reduce residual stresses) are recommended. H 2 control to reduce

  8. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  9. An assessment of BWR (boiling water reactor) Mark III containment challenges, failure modes, and potential improvements in performance

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1991-01-01

    This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs.

  10. Optimization of operation schemes in boiling water reactors using neural networks; Optimizacion de esquemas de operacion en reactores de agua en ebullicion usando redes neuronales

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Pelta, D. A., E-mail: juanjose.ortiz@inin.gob.mx [Universidad de Granada, Escuela Superior de Ingenierias, Informatica y Telecomunicacion, C/Daniel Saucedo Aranda s/n, 18071 Granada (Spain)

    2012-10-15

    In previous works were presented the results of a recurrent neural network to find the best combination of several groups of fuel cells, fuel load and control bars patterns. These solution groups to each problem of Fuel Management were previously optimized by diverse optimization techniques. The neural network chooses the partial solutions so the combination of them, correspond to a good configuration of the reactor according to a function objective. The values of the involved variables in this objective function are obtained through the simulation of the combination of partial solutions by means of Simulate-3. In the present work, a multilayer neural network that learned how to predict some results of Simulate-3 was used so was possible to substitute it in the objective function for the neural network and to accelerate the response time of the whole system of this way. The preliminary results shown in this work are encouraging to continue carrying out efforts in this sense and to improve the response quality of the system. (Author)

  11. Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

    International Nuclear Information System (INIS)

    Harper, J.R.; Garde, R.

    1981-11-01

    The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

  12. Power cycling experiments in INR-TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    Dumitru, M.

    2008-01-01

    The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

  13. Operating experiences at the Finnish TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    1988-01-01

    The Finnish TRIGA reactor has been in operation since March 1962. There are still 57 original Al-clad fuel elements in the core. So far we have had only two fuel cladding failures in 1981 and 1988. The first one was an Al-clad element and the second one a SS-clad. The low rate of fuel cladding failures has made it possible to use continuously also the Al-clad fuel elements. Although some conventional irradiations of certain type have been repeated successfully tens of times, new and unexpected incidents can still take place. As an example an event of a leaking irradiation capsule is described

  14. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

    Energy Technology Data Exchange (ETDEWEB)

    Fabreca, S. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-10-01

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

  15. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  16. Boiling nucleation

    International Nuclear Information System (INIS)

    Cole, R.

    1974-01-01

    Experimental results of flash evaporation of a pool of water subjected to sudden pressure drop are reported. The experiments were conducted with pure water at equilibrium temperatures between 40 to 80 0 C and with superheat in the range of about 3 to 5 0 C. Two distinct exponential decaying processes were identified for flash evaporation and the flashing time was found to decrease with an increase of equilibrium temperature and with the decrease of superheat. Basic experiments on flash evaporation of distilled water were conducted. However, the results may not be quantitatively applicable to seawater flash evaporators as the presence of salts in the seawater will considerably change the surface tension and in turn affect the nonequilibrium fraction

  17. Universality of oscillating boiling in Leidenfrost transition

    Science.gov (United States)

    Tran, Tuan; Khavari, Mohammad

    2017-11-01

    The Leidenfrost transition leads a boiling system to the boiling crisis, a state in which the liquid loses contact with the heated surface due to excessive vapor generation. Here, using experiments of liquid droplets boiling on a heated surface, we report a new phenomenon, termed oscillating boiling, at the Leidenfrost transition. We show that oscillating boiling results from the competition between two effects: separation of liquid from the heated surface due to localized boiling, and rewetting. We argue theoretically that the Leidenfrost transition can be predicted based on its link with the oscillating boiling phenomenon, and verify the prediction experimentally for various liquids. This work was funded by Nanyang Technological University and A*STAR, Singapore.

  18. Temperature distribution and local boiling behind a central blockage in a simulated FBR subassembly

    International Nuclear Information System (INIS)

    Brook, A.J.; Huber, F.; Peppler, W.

    1976-01-01

    A series of experiments has been carried out to investigate the effects of localised disturbance to the normal coolant flow in a fast reactor fuel element. The tests involved an electrically heated bundle of 169 pins, with a centrally located blockage extending over 49% of the flow area. Test section geometry corresponded to the SNR 300 Mk 1a fuel element. Measured temperature distributions behind the blockage agreed well with those measured in corresponding water experiments. The observed features of local boiling are discussed, and it is shown that a continued capability for cooling the blockage region is preserved, even with intensive local boiling

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  20. A study on boiling heat transfer with mixture boiling from vertical rod fin

    International Nuclear Information System (INIS)

    Kim, M.C.

    1981-01-01

    The purpose of the present study is concerned with the boiling characteristic of variations of the length-diameter ratio on the heat transfer rate where the nucleate boiling and natural convection occurred simultaneously. Circular fins were made with copper rod 32 mm in diameter, and those surfaces were mirror finished. The length-diameter ratio was varied 1 to 6. As a boiling liquid, the distilled water was used in this experiment. The results of this experiment were obtained as below. 1) From the observations, it was confirmed that nucleate boiling and natural convection occurred simultaneously. 2) As the length-diameter ratio increased, the boiling heat transfer rate also augmented. (author)

  1. Visualization of He II boiling process under the microgravity condition for 4.7 s by using a drop tower experiment

    Science.gov (United States)

    Takada, Suguru; Kimura, Nobuhiro; Pietrowicz, Sławomir; Grunt, Krzysztof; Murakami, Masahide; Okamura, Takahiro

    2018-01-01

    Superfluid helium (He II) has been utilized in space projects such as in the X-ray telescope, where it served as the heat sink of adiabatic demagnetization refrigerators. The study of He II boiling under microgravity might contribute to the construction of an important database facilitating the design of future space missions. Therefore, in this study, a visualization experiment of He II boiling was conducted under microgravity conditions by using the drop tower located at ZARM (Center of Applied Space Technology and Microgravity) in Bremen. The ZARM drop tower can provide up to 4.7 s of microgravity conditions in the utilized operation mode. The behavior of thermally induced bubbles during their growth and shrinkage was visualized using two high-speed cameras. A thin manganin wire was utilized as the heater. During the free fall period, the visualized bubble closely approached a steady state. The behavior can be roughly calculated using a simple equation based on kinetic theory.

  2. The MEDICI reactor cavity model: Chapter 5

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Trebilcock, W.

    1983-01-01

    The MEDICI reactor cavity model is currently under development with the goal of providing a flexible, relatively realistic treatment of ex-vessel severe accident phenomena suitable for large system codes like CONTAIN and MELCOR. The code is being developed with an emphasis on top-down design, to facilitate adaptability and multiple applications. A brief description of the overall code structure is provided. One of the key new models is then described in more detail. This is a dynamic quench model for debris beds. An example calculation using this model is presented. The question of whether it is necessary to consider the simultaneous motion of the quench front and ablation of the concrete is addressed with some scoping models. It is found that for realistic parameters and coolable beds, concrete ablation is too slow a process to be important on the quenching time scale. Remelt in the dry zone, however, is found to be potentially important on this time scale, so quench and remelt are considered simultaneously

  3. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    International Nuclear Information System (INIS)

    Chad Pope

    2007-01-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  4. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    2007-12-01

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  5. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)

  6. LEU-fuelled SLOWPOKE-2 research reactors: Operational experience and utilisation

    International Nuclear Information System (INIS)

    Kennedy, G.; St Pierre, J.; Bennett, L.G.I.; Nielsen, K.S.

    2002-01-01

    Atomic Energy of Canada Limited designed the SLOWPOKE-2 research reactor [1] based on experience with the SLOWPOKE-1 prototype, which operated for four years at the University of Toronto. Between 1976 and 1984, seven SLOWPOKE-2 reactors with HEU fuel were commissioned in six Canadian cities and in Kingston, Jamaica. They use 93% enriched uranium in the form of 28% uranium-aluminum alloy with aluminum cladding. The core is an assembly of about 300 fuel pins, only 22 cm diameter and 23 cm high, surrounded by a fixed beryllium annulus and a bottom beryllium slab. Criticality is maintained by adding beryllium plates in a tray on top of the core. The newer LEU fuel is constructed in the same manner as CANDU fuel, Zircaloy-clad, but with 20% enrichment instead of natural uranium oxide. The pool, reactor water container, light water moderator, beryllium reflector, cadmium control rod and irradiation sites are the same as for the HEU-fuelled reactors. A very important feature of the SLOWPOKE-2 reactor is its inherent safety due to its limited excess reactivity and large negative temperature coefficient of reactivity. With both the HEU and the LEU cores, power excursions were carried out during commissioning up to the maximum credible reactivity insertion of 4.3 mk. For both types of cores, the power reached about 80 kW after about 2 minutes before levelling off. At lower powers (lower temperatures), the negative feedback with temperature increase is greater with the HEU core and, for reactivity insertions less than 4.3 mk, lower powers are reached with the HEU core. Both cores are highly undermoderated but the HEU core even more so because there is less water in the volume of the core. At 80 kW some boiling may take place with both cores, causing rapid negative feedback, and the LEU core has additional negative feedback due to the U-238. Considering all the reactivity feedback effects, the HEU and LEU cores are roughly equivalent from a safety standpoint, since neither can

  7. RA research nuclear reactor operation in forced regime, Annex 5

    International Nuclear Information System (INIS)

    Mitrovic, S.

    1964-01-01

    In order to increase the flux and experimental potentials of the RA reactor, properties of the reactor operating in the forced regime have been inspected four times, twice with the spent fuel (core C), and twice with partly fresh fuel (cores E and F). Forced regime means power higher than nominal 6.5 MW. From the analyses of technology parameters during operation in the forced regime at power levels of 7 and 11 MW, state and behaviour of the components, efficiency of the biological protection and other parameters it was concluded that the reactor operation is possible at power levels of 10 MW during a whole year, and under certain conditions even longer operation at power levels up to 15 MW. This paper examines the reactor operation conditions in the forced regime and gives conclusions related to real increase of the experimental possibilities, isotope production and economic factors [sr

  8. Dry cooling tower operating experience in the LOFT reactor

    International Nuclear Information System (INIS)

    Hunter, J.A.

    1980-01-01

    A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

  9. Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

    OpenAIRE

    Algirdas Kaliatka; Eugenijus Uspuras; Sigitas Rimkevicius

    2007-01-01

    Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-ter...

  10. Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

    International Nuclear Information System (INIS)

    Yang Zhimin

    2000-01-01

    The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

  11. Radiological characterisation experience with Magnox reactors

    International Nuclear Information System (INIS)

    Westall, Bill A.; Towton, Barbara

    2012-01-01

    At the end of generation, power reactors will be decommissioned. Whether decommissioning is prompt or deferred, knowledge of the radioactive inventory of plant and structures is needed to develop and underpin the decommissioning strategy. As decommissioning progresses the level of detail required for the radioactive inventory increases as more specific and detailed questions need answering. Failure to adequately characterise will result in increased costs and project overruns due to missing optimal solutions, over pessimistic assumptions or unforeseen problems and regulatory issues. Radiological characterisation for decommissioning of Magnox power stations in the UK has been in progress for over a quarter of a century. Firstly measurements and calculations were carried out to develop a strategy. These have been followed by measurements to determine radioactive inventories of waste streams and packages or to allow decontamination of structures and most recently for partial de-licensing of sites. Some examples of the work carried out for the Magnox stations will be given, ranging from the neutron activation calculations to estimate the radioactive inventory within a bio-shield and measurements to validate them. Various plant and structures where the radioactive inventory is due to contamination have been characterised by measurements and examples for boilers and cooling ponds will be discussed. Various routine and ad-hoc measurements and shielding assessments have been performed on waste forms to help satisfy conditions for acceptance for disposal or exemption, which will be reviewed. Finally the measurements for de-licensing and the successful application of Data Quality Objectives will be addressed. (authors)

  12. Assessment of light water reactor accident management programs and experience

    International Nuclear Information System (INIS)

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation

  13. Fast reactor fuel development in Germany: Irradiation experience

    Energy Technology Data Exchange (ETDEWEB)

    Kummerer, K.R. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.). Inst. fuer Material- und Festkoerperforschung 3 - Teilinstitut Brennelemente); Muehling, G. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.). Projekt Schneller Brueter)

    1990-04-01

    Within the German Fast Breeder Project an extensive effort has been devoted to the development of fast reactor fuel elements, mostly in close cooperation between the Nuclear Research Center Karlsruhe and partners from industry and from other European ''breeder groups''. The main objective was the design and qualification of the envisaged reference fuel element with mixed oxide fuel and austenitic stainless steel cladding and structure. In this context a manifold irradiation programme in different European test reactors covered the normal standard operation conditions as well as above normal incidents and hypothetical accidents. The whole network of experiments resulted in sufficient experience for the design and realization of the prototype fast reactor power station SNR 300 in Kalkar. (orig.).

  14. Liquid jet experiments: relevance to inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1981-01-01

    In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

  15. An Overview of the International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gulliford, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  16. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  17. Experience, status and future of the computerized reactor instrumentation at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Frankl, M.; Boeck, H.; Katrik, P.; Schachner, H.

    1997-01-01

    The paper describes the 33 years old history of the instrumentation of the TRIGA reactor Vienna and focuses on the present computerized instrumentation installed in 1992. The experience of three years of operation is discussed and some of the failures are analyzed. Potential future problems both with soft- and hardware as well as with spare part supplies are analyzed. (author). 6 figs

  18. PROSPECT: A Short-baseline Reactor Precision Spectrum and Oscillation Experiment

    Science.gov (United States)

    Langford, Thomas; Prospect Collaboration

    2015-10-01

    PROSPECT is a phased experiment consisting of segmented Li-loaded liquid scintillator antineutrino detectors designed to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. The experiment will be located at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. The first phase is a movable 2.5 tonne detector located 7-9 m from the compact, highly enriched uranium (HEU) core. Over the past two years, PROSPECT has deployed multiple prototype detectors at HFIR to understand the local background environment and demonstrate active and passive background rejection. Measuring the neutrino spectrum from 235U will give insight to the recent spectral discrepancies and provide an important benchmark for future reactor experiments. As a high statistics experiment, PROSPECT will probe the sterile neutrino best-fit region within one year of operation at HFIR.

  19. Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

    International Nuclear Information System (INIS)

    Bowman, S.M.; Suto, T.

    1996-10-01

    ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

  20. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  1. Some critical corrosion issues and mitigation strategies affecting light water reactors

    International Nuclear Information System (INIS)

    Jones, R.L.

    1996-01-01

    Recent corrosion experience in US light water reactor nuclear power plants is reviewed with emphasis on mitigation strategies to control the cost of corrosion to light water reactor operators. The most costly issues have been stress corrosion cracking of stainless steel coolant piping in boiling water reactors and corrosion damage to steam generator tubes in pressurized water reactors. Other significant corrosion problems for both reactor types are briefly reviewed

  2. Review of experiments for research reactors - approved 1974

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This standard establishes guidelines for the review and approval of experiments performed at research reactor facilities. This standard identifies the major areas that shall be reviewed for each experiment to ensure that it (a) falls within the limits delineated in the technical specifications, (b) does not present an unreviewed safety question as defined in 10 CFR Section 50.59 π2-, (c) does not constitute a threat to the health and safety of any individual or group of individuals, and (d) does not constitute a hazard to the reactor facility or other equipment. In addition, this standard recommends a system for classifying experiments to establish levels of review and approval commensurate with the level of risk inherent in the experiment

  3. Heavy reflector experiments in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Silva, Graciete Simoes de Andrade e; Mura, Luis Felipe; Fuga, Rinaldo; Jerez, Rogerio; Mendonca, Arlindo Gilson

    2012-01-01

    Full text: The heavy reflector experiments performed in the IPEN/MB-01 research reactor facility comprise a set of critical configurations employing the standard 28x26-fuel-rod configuration. The heavy reflector either Stainless Steel, Carbon Steel or Nickel plates was placed at one of the faces of the IPEN/MB-01 reactor. Criticality is achieved by inserting the control banks BC1 and BC2 to the critical position. 32 plates around 0.3 mm thick were used in the experiment. The chosen distance between last fuel rod row and the first laminate for both type of laminates was 5.5 mm. Considering initially the SS case, the experimental data reveal that the reactivity decreases up to the sixth plate and after that it increases, becomes nearly zero (which was equivalent to initial zero excess reactivity with zero plates) for the 21 plates case and reaches a value of 154.91 pcm when the whole set of 32 plates are inserted in the reflector. This is a very striking result because it demonstrates that when all 32 plates are inserted in the reflector there is a net gain of reactivity. The reactivity behavior demonstrates all the physics events already mentioned in this work. When the number of plates are small (around 6), the neutron absorption in the plates is more important than the neutron reflection and the reactivity decreases. This condition holds up to a point where the neutron reflection becomes more important than the neutron absorption in the plates and the reactivity increases. The experimental data for the Carbon Steel and Nickel case shows the main features of the SS case, but for the Carbon Steel case the reactivity gain is small, thus demonstrating that Carbon Steel or essentially iron has not the reflector capability as the SS laminates do. The measured data of Nickel plates show a higher reactivity gain, thus demonstrating that Nickel is a good reflector. The theoretical analysis employing MCNP5 and ENDF/B-VII.0 show that the SS calculated results are in a good

  4. Fission product behavior in the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

    1975-10-01

    Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

  5. Nuclear research reactor 0.5 to 3 MW

    International Nuclear Information System (INIS)

    1992-05-01

    This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

  6. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  7. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

  8. Simulation experiments for radiologic inspection inside a nuclear reactor

    Science.gov (United States)

    Wei, W.; Murphy, R. V.; Sonnenburg, D. K.

    1996-02-01

    Experiments were performed in a gamma cell to simulate radiologic inspection inside a CANDU nuclear power reactor. Radiation in the gamma cell is similar, both in magnitude and directions, to that in a shut down CANDU reactor. The inspection consists of detecting garter spring spacers used to maintain the gap between pressure tubes and calandria tubes in CANDU reactors. A shielding head made of tungsten alloy was placed inside a pressure tube. A glass scintillator was used as the gamma radiation detector. The scintillation light travelled through a fiber-optic light guide to the radiation-free environment outside the gamma cell, where the light was detected by a light sensor. Earlier experiments used a high-resolution CCD (charge coupled device) camera to capture images transferred through a high-resolution fiberscope. Later experiments used silicon photodiodes to measure the intensity of the light transferred through a fiber bundle that was made in-house. The light intensity approach was found to be more suitable for the detection of garter springs. The removal of the garter spring resulted in an immediate increase in the intensity of the scintillation light. Experimental results in the gamma cell show great promise for constructing a real-time garter spring detection tool for use in CANDU reactors.

  9. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    Energy Technology Data Exchange (ETDEWEB)

    Briere, E.; Larrauri, D.; Olive, J. [Electricite de France, Chatou (France)

    1995-09-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu`s criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF`s program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part.

  10. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    International Nuclear Information System (INIS)

    Briere, E.; Larrauri, D.; Olive, J.

    1995-01-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu's criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF's program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part

  11. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  12. Latest Results from the Daya Bay Reactor Neutrino Experiment

    CERN Multimedia

    CERN. Geneva

    2014-01-01

    Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

  13. Nuclear heating of materials in the R-5 reactor

    International Nuclear Information System (INIS)

    Khosla, S.K.; Sankaranarayanan, S.

    1976-01-01

    For the safety evaluation of irradiation assemblies in reactors, the estimation of energy deposition rates due to pile neutron and gamma radiations, in constructional materials is essential. The calculational methods used for the R-5 reactor to obtain the energy deposition rates due to fast neutron scattering and gamma ray absorption in heavy water are summarised. For other materials of interest suitable scaling factors are used based on simple physics arguments. Typical values for materials like aluminium, copper, lead, light water, stainless steel and zircaloy are indicated. Applications of the results to typical in-core experimental assemblies are discussed. (author)

  14. Nuclear data and integral experiments in reactor physics

    International Nuclear Information System (INIS)

    Farinelli, U.

    1980-01-01

    The material given here broadly covers the content of the 10 lectures delivered at the Winter Course on Reactor Theory and Power Reactors, ICTP, Trieste (13 February - 10 March 1978). However, the parts that could easily be found in the current literature have been omitted and replaced with the appropriate references. The needs for reactor physics calculations, particularly as applicable to commercial reactors, are reviewed in the introduction. The relative merits and shortcomings of fundamental and semi-empirical methods are discussed. The relative importance of different nuclear data, the ways in which they can be measured or calculated, and the sources of information on measured and evaluated data are briefly reviewed. The various approaches to the condensation of nuclear data to multigroup cross sections are described. After some consideration to the sensitivity calculations and the evaluation of errors, some of the most important type of integral experiments in reactor physics are introduced, with a view to showing the main difficulties in the interpretation and utilization of their results and the most recent trends in experimentation. The conclusions try to assign some priorities in the implementation of experimental and calculational capabilities, especially for a developing country. (author)

  15. Training experience at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report

  16. Recent experience reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    Vignes, A.

    1985-01-01

    This paper present the Framatome's recent experience in the manufacture of 1300 MWe PWR vessels; one shows how the very high standards of quality have been obtained to meet the stringent requirements. After a description of a pressure vessel, materials and forgings properties are presented. The nature and sequence of the main fabrication operations are reviewed. This paper deals after with the quality of welds, the preheating and post-heating equipment, the submerged arc welding process and procedures, the cladding process, and the under-clad cracking problems. Ultrasonic inspection procedures of the main welds are described with a comparison of RCCM (design and construction rules for mechanical components of PWR units) and Sizewell B specifications. Support of data on the reproductibility and effectiveness of ultrasonic examination and on the reliability given by repetitive inspection are presented

  17. Innovative nuclear technologies based on radiation induced surface activation. (5) Development of high performance BWR by the radiation induced surface activation visualization study on the boiling enhancement with irradiation

    International Nuclear Information System (INIS)

    Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

    2004-01-01

    Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, CHF of metal oxides irradiated by gamma rays were investigated. The heating test section made of titanium was 0.5 mm in diameter. Oxidation of the surface was carried out by plasma jetting. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. A test piece had been hold horizontally on the electrode after 5400 kGy irradiation. Then, the whole CHF test apparatus with test piece was set on the table in the gamma ray irradiation room. The test piece was irradiated in the water at least 30 minutes. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure under irradiation. The results of on-site experiment were compared with that of off-site one. (author)

  18. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  19. Chasing {theta}{sub 13} with new reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

    2005-12-15

    It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the {theta}{sub 13} mixing angle, free from any parameter degeneracies and correlations induced by matter effect and the unknown leptonic Dirac CP phase. The current best constraint on the third mixing angle comes from the Chooz reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90 % C.L., {delta}m{sub atm}{sup 2}=2.010{sup -3} eV{sup 2}). Several projects of experiment, with different timescales, have been proposed over the last two years all around the world. Their sensitivities range from sin{sup 2}(2{theta}{sub 13})<0.01 to 0.03, having thus an excellent discovery potential of the {nu}{sub e} fraction of {nu}{sub 3}.

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

    Science.gov (United States)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  2. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

    International Nuclear Information System (INIS)

    1963-01-01

    At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  4. Study for Reactor Monitoring using Anti-neutrino Detection in the Neos experiment

    Energy Technology Data Exchange (ETDEWEB)

    Han, Bo Young; Sun, Gwang Min [KAERI, Daejeon (Korea, Republic of); Jeon, Eun Ju [ISB, Daejeon (Korea, Republic of); and others

    2016-05-15

    In this study we describe a feasibility study of reactor monitoring using antineutrino detection in the Neutrino Experiment for Oscillation at Short baseline (NEOS) at Hanbit power plant. Recently, in the perspective of nonproliferation issues and misuse of nuclear energy as a fast-growing nuclear energy industry, the application of anti-neutrino measurement has been proposed and the feasibility studies has been carried out as a novel technology for monitoring the burning process of nuclear power reactor. The NEOS detector with 1000 L Gd-doped liquid scintillator was installed in tendon gallery at Hanbit power station unit 5 and has been collecting close to 2000 IBD events per day with the signal to noise ratio of ∼ 20. As a preliminary result, we demonstrate the possibility of monitoring nuclear power reactor with the IBD counting rate during reactor power ON, ramping up, and OFF.

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

    International Nuclear Information System (INIS)

    Katto, Yoshiro; Ishii, Kazunori.

    1978-01-01

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

  6. Boiling Point

    Science.gov (United States)

    Jansen, Michael C.

    2002-01-01

    The author recounts his experiences he helped to investigate the accident which destroyed the Space Shuttle Challenger. The focus was on how he used novel approaches to investigate heat transfer in the shuttle's hydrogen tank, after an expert he sought for advice proved unhelpful.

  7. Reactor D and D at Argonne National Laboratory - lessons learned

    International Nuclear Information System (INIS)

    Fellhauer, C. R.

    1998-01-01

    This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

  8. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: fundamental aspects of two-phase flow and boiling heat transfer

    International Nuclear Information System (INIS)

    1980-08-01

    Separate abstracts are included for each of the papers presented concerning critical flow of two-phase mixtures; two-phase flow instrumentation; critical heat flux and effects of local disturbances; heat transfer and rewetting during reflood; hydrodynamic mechanisms in boiling heat transfer; and entrainment and droplet deposition in two-phase flow. Five papers have been previously abstracted and input to the data base

  9. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    Stankovskiy, Evgeny Yuryevich

    data and simulated responses of neutron detectors. The accuracy (0.2% uncertainty) of the calculated effective delayed neutron fraction, together with the exponential decay of neutron population in the reactor, allows the estimation of the mean neutron generation time to be performed with acceptable uncertainty (1.5%). Because the multiplication constant is a standard result with MCNP, the difference between dynamic reactivity (which is measured in the experiment) and static reactivity is clearly shown.

  10. Numerical experiments on evaporation and explosive boiling of ultra-thin liquid argon film on aluminum nanostructure substrate.

    Science.gov (United States)

    Wang, Weidong; Zhang, Haiyan; Tian, Conghui; Meng, Xiaojie

    2015-01-01

    Evaporation and explosive boiling of ultra-thin liquid film are of great significant fundamental importance for both science and engineering applications. The evaporation and explosive boiling of ultra-thin liquid film absorbed on an aluminum nanostructure solid wall are investigated by means of molecular dynamics simulations. The simulated system consists of three regions: liquid argon, vapor argon, and an aluminum substrate decorated with nanostructures of different heights. Those simulations begin with an initial configuration for the complex liquid-vapor-solid system, followed by an equilibrating system at 90 K, and conclude with two different jump temperatures, including 150 and 310 K which are far beyond the critical temperature. The space and time dependences of temperature, pressure, density number, and net evaporation rate are monitored to investigate the phase transition process on a flat surface with and without nanostructures. The simulation results reveal that the nanostructures are of great help to raise the heat transfer efficiency and that evaporation rate increases with the nanostructures' height in a certain range.

  11. Development of an experimental apparatus for boiling analysis

    International Nuclear Information System (INIS)

    Castro, A.J.A. de.

    1984-04-01

    The nucleate boiling is the most interesting boiling regime for practical appliccations, including nuclear reactor engineering. such regime is characterized by very high heat transfer rates with only small surface superheating. An experimental apparatus is developed for studying parameters which affect nucleate boiling. The following parameters are analysed: pressure, fluid velocity and the fluid temperature at the test section entrance. The performance of experimental apparatus is analysed by results and by problems raised by the oeration of setup. (Author) [pt

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

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Gao, C.

    1996-01-01

    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 (q MHF ) was highest at the lowermost position and decreased with increased local inclination on the surface. Both local and surface average q MHF 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 q MHF 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 q MHF by as much as 40%

  13. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1986-05-01

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

  14. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    1986-01-01

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  15. Pool Boiling of Hydrocarbon Mixtures on Water

    Energy Technology Data Exchange (ETDEWEB)

    Boee, R.

    1996-09-01

    In maritime transport of liquefied natural gas (LNG) there is a risk of spilling cryogenic liquid onto water. The present doctoral thesis discusses transient boiling experiments in which liquid hydrocarbons were poured onto water and left to boil off. Composition changes during boiling are believed to be connected with the initiation of rapid phase transition in LNG spilled on water. 64 experimental runs were carried out, 14 using pure liquid methane, 36 using methane-ethane, and 14 using methane-propane binary mixtures of different composition. The water surface was open to the atmosphere and covered an area of 200 cm{sup 2} at 25 - 40{sup o}C. The heat flux was obtained by monitoring the change of mass vs time. The void fraction in the boiling layer was measured with a gamma densitometer, and a method for adapting this measurement concept to the case of a boiling cryogenic liquid mixture is suggested. Significant differences in the boil-off characteristics between pure methane and binary mixtures revealed by previous studies are confirmed. Pure methane is in film boiling, whereas the mixtures appear to enter the transitional boiling regime with only small amounts of the second component added. The results indicate that the common assumption that LNG will be in film boiling on water because of the high temperature difference, may be questioned. Comparison with previous work shows that at this small scale the results are influenced by the experimental apparatus and procedures. 66 refs., 76 figs., 28 tabs.

  16. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  17. Present and Future Experiments in Non-equilibrium Reactor Antineutrino Energy Spectrum

    OpenAIRE

    Kopeikin, V. I.; Mikaelyan, L. A.

    2005-01-01

    Considerable efforts that have been undertaken in the recent years in low energy antineutrino experiments require further systematic investigations in line of reactor antineutrino spectroscopy as a metrological basis of these experiments. We consider some effects associated with the non-equilibrium of reactor antineutrino radiation and residual antineutrino emission from spent reactor fuel in contemporary antineutrino experiments.

  18. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  19. Irradiation effects test series, test IE-5. Test results report. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Croucher, D. W.; Yackle, T. R.; Allison, C. M.; Ploger, S. A.

    1978-01-01

    Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod behavior during film boiling operation. The four rods were subjected to a preconditioning period, a power ramp to an average fuel rod peak power of 65 kW/m, and steady state operation for one hour at a coolant mass flux of 4880 kg/s-m/sup 2/ for each rod. After a flow reduction to 1800 kg/s-m/sup 2/, film boiling occurred on one rod. Additional flow reductions to 970 kg/s-m/sup 2/ produced film boiling on the three remaining fuel rods. Maximum time in film boiling was 80s. The rod having the highest initial internal pressure (8.3 MPa) failed 10s after the onset of film boiling. A second rod failed about 90s after reactor shutdown. The report contains a description of the experiment, the test conduct, test results, and results from the preliminary postirradiation examination. Calculations using a transient fuel rod behavior code are compared with the test results.

  20. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning